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A molten salt reactor (MSR) is a class of nuclear fission reactor in which the primary nuclear reactor coolant and/or the fuel is a molten salt mixture. A key characteristic of MSRs is their operation at or close to atmospheric pressure, rather than the 75-150 times atmospheric pressure of typical light-water reactors (LWR), hence reducing the large, expensive containment structures used for LWRs and eliminating a source of explosion risk. Another key characteristic of MSRs is higher operating temperatures than a traditional LWR, providing higher electricity-generation efficiency and, in some cases, process-heat opportunities. Relevant design challenges include the corrosivity of hot salts and the changing chemical composition of the salt as it is transmuted by reactor radiation. MSR cost estimations are uncertain but comparable or cheaper than LWRs.
While many design variants have been proposed, there are three main categories regarding the role of molten salt:
|Molten salt fuel - circulating||ARE • MSRE • DMSR • MSFR • LFTR • IMSR • AWB, CMSR • EVOL • DFR • TMSR-500|
|Molten salt fuel - static||SSR|
|Molten salt coolant only||TMSR • FHR|
(The use of molten salt as fuel and as coolant are independent design choices - the original circulating-fuel-salt MSRE and the more recent static-fuel-salt SSR use salt as fuel and salt as coolant; the DFR uses salt as fuel but metal as coolant; and the FHR has solid fuel but salt as coolant)
MSRs offer multiple advantages over conventional nuclear power plants, although for historical reasonsthey have not been deployed.
The concept was first established in the 1950s. The early Aircraft Reactor Experiment was primarily motivated by the compact size that the technique offers, while the Molten-Salt Reactor Experiment aimed to prove the concept of a nuclear power plant which implements a thorium fuel cycle in a breeder reactor. Increased research into Generation IV reactor designs began to renew interest in the technology.
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MSR research started with the U.S. Aircraft Reactor Experiment (ARE) in support of the U.S. Aircraft Nuclear Propulsion program. ARE was a 2.5 MWth nuclear reactor experiment designed to attain a high energy density for use as an engine in a nuclear-powered bomber.
The project included experiments, including high temperature and engine tests collectively called the Heat Transfer Reactor Experiments: HTRE-1, HTRE-2 and HTRE-3 at the National Reactor Test Station (now Idaho National Laboratory) as well as an experimental high-temperature molten salt reactor at Oak Ridge National Laboratory – the ARE.
ARE used molten fluoride salt NaF-ZrF4-UF4 (53-41-6 mol%) as fuel, moderated by beryllium oxide (BeO). Liquid sodium was a secondary coolant.
The experiment had a peak temperature of 860 °C. It produced 100 MWh over nine days in 1954. This experiment used Inconel 600 alloy for the metal structure and piping.
An MSR was operated at the Critical Experiments Facility of the Oak Ridge National Laboratory in 1957. It was part of the circulating-fuel reactor program of the Pratt & Whitney Aircraft Company (PWAC). This was called Pratt and Whitney Aircraft Reactor-1 (PWAR-1). The experiment was run for a few weeks and at essentially zero power, although it reached criticality. The operating temperature was held constant at approximately 675 °C (1,250 °F). The PWAR-1 used NaF-ZrF4-UF4 as the primary fuel and coolant. It was one of three critical MSRs ever built.
Oak Ridge National Laboratory (ORNL) took the lead in researching MSRs through the 1960s. Much of their work culminated with the Molten-Salt Reactor Experiment (MSRE). MSRE was a 7.4 MWth test reactor simulating the neutronic "kernel" of a type of epithermal thorium molten salt breeder reactor called the liquid fluoride thorium reactor (LFTR). The large (expensive) breeding blanket of thorium salt was omitted in favor of neutron measurements.
MSRE's piping, core vat and structural components were made from Hastelloy-N, moderated by pyrolytic graphite. It went critical in 1965 and ran for four years. Its fuel was LiF-BeF2-ZrF4-UF4 (65-29-5-1). The graphite core moderated it. Its secondary coolant was FLiBe (2LiF-BeF2). It reached temperatures as high as 650 °C and achieved the equivalent of about 1.5 years of full power operation.
The culmination of the ORNL research during the 1970–1976 timeframe resulted in a molten salt breeder reactor (MSBR) design. Fuel was to be LiF-BeF2-ThF4-UF4 (72-16-12-0.4) with graphite moderator. The secondary coolant was to be NaF-NaBF4. Its peak operating temperature was to be 705 °C. It would follow a 4-year replacement schedule. The MSR program closed down in the early 1970s in favor of the liquid metal fast-breeder reactor (LMFBR), after which research stagnated in the United States. As of 2011 [update] , ARE and MSRE remained the only molten-salt reactors ever operated.
The MSBR project received funding from 1968 to 1976 of (in 2019 dollars ) $66.4 million.
Officially, the program was cancelled because:
Engel et al 1980 "examine the conceptual feasibility of a molten-salt power reactor fueled with denatured uranium-235 (i.e. with low-enriched uranium) and operated with a minimum of chemical processing." The main priority behind the design characteristics was proliferation resistance.Although the DMSR can theoretically be fueled partially by thorium or plutonium, fueling solely with low enriched uranium (LEU) helps maximize proliferation resistance.
Other important goals of the DMSR were to minimize R&D and to maximize feasibility. The Generation IV international Forum (GIF) includes "salt processing" as a technology gap for molten salt reactors.The DMSR requires minimal chemical processing because it is a burner rather than a breeder. Both reactors built at ORNL were burner designs. In addition, the choices to use graphite for neutron moderation and enhanced Hastelloy-N for piping simplified the design and reduced R&D.
The UK's Atomic Energy Research Establishment (AERE) were developing an alternative MSR design across its National Laboratories at Harwell, Culham, Risley and Winfrith. AERE opted to focus on a lead-cooled 2.5 GWe Molten Salt Fast Reactor (MSFR) concept using a chloride. They also researched helium gas as a coolant.
The UK MSFR would be fuelled by plutonium, a fuel considered to be 'free' by the program's research scientists, because of the UK's plutonium stockpile.
Despite their different designs, ORNL and AERE maintained contact during this period with information exchange and expert visits. Theoretical work on the concept was conducted between 1964 and 1966, while experimental work was ongoing between 1968 and 1973. The program received annual government funding of around £100,000–£200,000 (equivalent to £2m–£3m in 2005). This funding came to an end in 1974, partly due to the success of the Prototype Fast Reactor at Dounreay which was considered a priority for funding as it went critical in the same year.
In the USSR, a molten-salt reactor research program was started in the second half of the 1970s at the Kurchatov Institute. It included theoretical and experimental studies, particularly the investigation of mechanical, corrosion and radiation properties of the molten salt container materials. The main findings supported the conclusion that no physical nor technological obstacles prevented the practical implementation of MSRs.
MSR interest resumed in the new millennium with continuing delays in fusion power and other nuclear power programs.
The LFTR design was strongly supported by Alvin Weinberg, who patented the light-water reactor and was a director of the U.S.'s Oak Ridge National Laboratory. In 2016 Nobel prize winning physicist Carlo Rubbia, former Director General of CERN, claimed that one of the main reasons why research was cut is that thorium is difficult to turn into a nuclear weapon.
Terrestrial Energy, a Canadian-based company, is developing a DMSR design called the Integral Molten Salt Reactor (IMSR). The IMSR is designed to be deployable as a small modular reactor (SMR). Their design currently undergoing licensing is 400MW thermal (190MW electrical). With high operating temperatures, the IMSR has applications in industrial heat markets as well as traditional power markets. The main design features include neutron moderation from graphite, fueling with low-enriched uranium and a compact and replaceable Core-unit. Decay heat is removed passively using nitrogen (with air as an emergency alternative). The latter feature permits the operational simplicity necessary for industrial deployment.
Terrestrial completed the first phase of a prelicensing review by the Canadian Nuclear Safety Commission in 2017, which provided a regulatory opinion that the design features are generally safe enough to eventually obtain a license to construct the reactor.
China initiated a thorium molten-salt reactor research project in January 2011.A 100 MW demonstrator of the solid fuel version (TMSR-SF), based on pebble bed technology, planned to be ready by 2024. Initially, a 10 MW pilot and a larger demonstrator of the liquid fuel (TMSR-LF) variant were targeted for 2024 and 2035, respectively. China then accelerated its program to build two 12 MW reactors underground at Wuwei research facilities in Gansu Province by 2020, beginning with the TMSR-LF1 prototype. Heat from the thorium molten-salt reaction would be used to produce electricity, hydrogen, industrial chemicals, desalination and minerals. The project also seeks to test new corrosion-resistant materials.
In 2017, ANSTO/Shanghai Institute Of Applied Physics announced the creation of a NiMo-SiC alloy for use in MSRs.
Copenhagen Atomics is a Danish molten salt technology company developing mass manufacturable molten salt reactors. The Copenhagen Atomics Waste Burner is a single-fluid, heavy water moderated, fluoride-based, thermal spectrum and autonomously controlled molten salt reactor. This is designed to fit inside of a leak-tight, 40-foot, stainless steel shipping container. The heavy water moderator is thermally insulated from the salt and continuously drained and cooled to below 50 °C. A molten lithium-7 deuteroxide (7LiOD) moderator version is also being researched. The reactor utilizes the thorium fuel cycle using separated plutonium from spent nuclear fuel as the initial fissile load for the first generation of reactors, eventually transitioning to a thorium breeder. Copenhagen Atomics is actively developing and testing valves, pumps, heat exchangers, measurement systems, salt chemistry and purification systems, and control systems and software for molten salt applications.
Seaborg Technologies is developing the core for a Compact Molten Salt Reactor (CMSR). The CMSR is a high temperature, single salt, thermal MSR designed to go critical on commercially available low enriched uranium. The CMSR design is modular, and uses proprietary NaOH moderator.The reactor core is estimated to be replaced every 12 years. During operation, the fuel will not be replaced and will burn for the entire 12-year reactor lifetime. The first version of the Seaborg core is planned to produce 250 MWth power and 100 MWe power. As a power plant, the CMSR will be able to deliver electricity, clean water and heating/cooling to around 200,000 households.
The CNRS project EVOL (Evaluation and viability of liquid fuel fast reactor system) project, with the objective of proposing a design of the MSFR (Molten Salt Fast Reactor),released its final report in 2014. R& Various MSR projects like FHR, MOSART, MSFR, and TMSR have common R&D themes.
The EVOL project will be continued by the EU-funded Safety Assessment of the Molten Salt Fast Reactor (SAMOFAR) project, in which several European research institutes and universities collaborate.
The German Institute for Solid State Nuclear Physics in Berlin has proposed the Dual fluid reactor as a concept for a fast breeder lead-cooled MSR. The original MSR concept used the fluid salt to provide the fission materials and also to remove the heat. Thus it had problems with the needed flowspeed. Using 2 different fluids in separate circles solves the problem.[ citation needed ]
In 2015, Indian researchers published a MSR design,as an alternative path to thorium-based reactors, according to India's three-stage nuclear power programme.
Thorcon is developing the TMSR-500 molten salt reactor for the Indonesian market.
The Fuji Molten Salt Reactor is a 100 to 200 MWe LFTR, using technology similar to the Oak Ridge project. A consortium including members from Japan, the U.S. and Russia are developing the project. The project would likely take 20 years to develop a full size reactor,but the project seems to lack funding.
The Russian MBIR is a planned 150 MWt, sodium-cooled fast reactor. It is to be a multi-loop research reactor for testing lead, lead-bismuth and gas coolants, with a MOX (mixed uranium and plutonium oxide) fuel. An on-site, pyrochemical, closed fuel cycle facility is planned. The reactor is planned to begin operation in 2020. As planned, it wil be the world's most-powerful research reactor.
The Alvin Weinberg Foundation is a British non-profit organization founded in 2011, dedicated to raising awareness about the potential of thorium energy and LFTR. It was formally launched at the House of Lords on 8 September 2011.It is named after American nuclear physicist Alvin M. Weinberg, who pioneered thorium MSR research.
The Stable Salt Reactor, designed by Moltex Energy, was selected as the most suitable of six MSR designs for UK implementation in a 2015 study commissioned by the UK's innovation agency, Innovate UK.UK government support has been weak, but Moltex has obtained support from New Brunswick Power for the development of a pilot plant in Point Lepreau, Canada, and financial backing from IDOM (an international engineering firm) and is currently engaged in the Canadian Vendor Design Review process.
Idaho National Laboratory designed a molten-salt-cooled, molten-salt-fuelled reactor with a prospective output of 1000 MWe.
Kirk Sorensen, former NASA scientist and chief nuclear technologist at Teledyne Brown Engineering, is a long-time promoter of the thorium fuel cycle, coining the term liquid fluoride thorium reactor. In 2011, Sorensen founded Flibe Energy, a company aimed at developing 20–50 MW LFTR reactor designs to power military bases. (It is easier to approve novel military designs than civilian power station designs in the US nuclear regulatory environment).
Transatomic Power pursued what it termed a Waste-Annihilating Molten Salt Reactor (acronym WAMSR), intended to consume existing spent nuclear fuel,from 2011 until ceasing operation in 2018.
In January 2016, the United States Department of Energy announced a $80m award fund to develop Generation IV reactor designs.One of the two beneficiaries, Southern Company will use the funding to develop a Molten Chloride Fast Reactor (MCFR), a type of MSR developed earlier by British scientists.
Nuclear reactors can be categorized in various ways. MSR designs participate in many of those categories. MSRs can be burners or breeders. They can be fast or thermal or epithermal.Thermal reactors typically employ a moderator (usually graphite) to slow the neutrons down and moderate temperature. They can accept a variety of fuels (low-enriched uranium, thorium, depleted uranium, waste products) and coolants (fluoride, chloride, lithium, beryllium, mixed). Fuel cycle can be either closed or once-through. They can be monolithic or modular, large or small. The reactor can adopt a loop, modular or integral configuration. Variations include:
(Also referred to as "fluoride salt-cooled high-temperature reactor" (FHR).)
This approach involves using a fluoride-salt as the coolant. Both the traditional MSR and the very-high-temperature reactor (VHTR) were selected as potential designs for study under the Generation Four Initiative (GEN-IV). One version of the VHTR under study was the Liquid-Salt Very-High-Temperature Reactor (LS-VHTR), also commonly called the Advanced High-Temperature Reactor (AHTR).[ citation needed ]
It uses liquid salt as a coolant in the primary loop, rather than a single helium loop. It relies on "TRISO" fuel dispersed in graphite. Early AHTR research focused on graphite in the form of graphite rods that would be inserted in hexagonal moderating graphite blocks, but current studies focus primarily on pebble-type fuel.[ citation needed ] The LS-VHTR can work at very high temperatures (the boiling point of most molten salt candidates is >1400 °C); low-pressure cooling that can be used to match hydrogen production facility conditions (most thermochemical cycles require temperatures in excess of 750 °C); better electric conversion efficiency than a helium-cooled VHTR operating in similar conditions; passive safety systems and better retention of fission products in the event of an accident.[ citation needed ]
Reactors containing molten thorium salt, called liquid fluoride thorium reactors (LFTR), would tap the thorium fuel cycle. Private companies from Japan, Russia, Australia and the United States, and the Chinese government, have expressed interest in developing this technology.
Advocates estimate that five hundred metric tons of thorium could supply U.S. energy needs for one year.The U.S. Geological Survey estimates that the largest-known U.S. thorium deposit, the Lemhi Pass district on the Montana-Idaho border, contains thorium reserves of 64,000 metric tons.
Traditionally, these reactors were known as Molten Salt Breeder Reactors (MSBRs) or Thorium Molten Salt Reactors (TMSRs), but the name LFTR was promoted as a rebrand in the early 2000s by Kirk Sorensen.
The Stable Salt Reactor is a relatively recent concept which holds the molten salt fuel statically in traditional LWR fuel pins. Pumping of the fuel salt, and all the corrosion/deposition/maintenance/containment issues arising from circulating a highly radioactive, hot and chemically complex fluid, are no longer required. The fuel pins are immersed in a separate, non-fissionable fluoride salt which acts as primary coolant.
MSRs can be cooled in various ways, including using molten salts.
Molten-salt-cooled solid-fuel reactors are variously called "molten salt reactor system" in the Generation IV proposal, Molten Salt Converter Reactors (MSCR), advanced high-temperature reactors (AHTRs), or fluoride high-temperature reactors (FHR, preferred DOE designation).
FHRs cannot reprocess fuel easily and have fuel rods that need to be fabricated and validated, requiring up to twenty years[ citation needed ] from project inception. FHR retains the safety and cost advantages of a low-pressure, high-temperature coolant, also shared by liquid metal cooled reactors. Notably, steam is not created in the core (as is present in BWRs), and no large, expensive steel pressure vessel (as required for PWRs). Since it can operate at high temperatures, the conversion of the heat to electricity can use an efficient, lightweight Brayton cycle gas turbine.
Much of the current research on FHRs is focused on small, compact heat exchangers that reduce molten salt volumes and associated costs.
Molten salts can be highly corrosive and corrosivity increases with temperature. For the primary cooling loop, a material is needed that can withstand corrosion at high temperatures and intense radiation. Experiments show that Hastelloy-N and similar alloys are suited to these tasks at operating temperatures up to about 700 °C. However, operating experience is limited. Still higher operating temperatures are desirable—at 850 °C thermochemical production of hydrogen becomes possible. Materials for this temperature range have not been validated, though carbon composites, molybdenum alloys (e.g. TZM), carbides, and refractory metal based or ODS alloys might be feasible.
A workaround suggested by a private researcher is to use the new beta-titanium Au alloys as this would also allow extreme temperature operation as well as increasing the safety margin.[ citation needed ]
A prototypical example of a dual fluid reactor is the lead-cooled, salt-fueled reactor.
The salt mixtures are chosen to make the reactor safer and more practical.
Fluorine has only one stable isotope (F-19), and does not easily become radioactive under neutron bombardment. Compared to chlorine and other halides, fluorine also absorbs fewer neutrons and slows ("moderates") neutrons better. Low-valence fluorides boil at high temperatures, though many pentafluorides and hexafluorides boil at low temperatures. They must be very hot before they break down into their constituent elements. Such molten salts are "chemically stable" when maintained well below their boiling points. Fluoride salts dissolve poorly in water, and do not form burnable hydrogen.
Chlorine has two stable isotopes (35
Cl and 37
Cl), as well as a slow-decaying isotope between them which facilitates neutron absorption by 35
Chlorides permit fast breeder reactors to be constructed. Much less research has been done on reactor designs using chloride salts. Chlorine, unlike fluorine, must be purified to isolate the heavier stable isotope, chlorine-37, thus reducing production of sulfur tetrachloride that occurs when chlorine-35 absorbs a neutron to become chlorine-36, then degrades by beta decay to sulfur-36.
Lithium must be in the form of purified 7
Li, because 6
Li effectively captures neutrons and produces tritium. Even if pure 7Li is used, salts containing lithium cause significant tritium production, comparable with heavy water reactors.
Reactor salts are usually close to eutectic mixtures to reduce their melting point. A low melting point simplifies melting the salt at startup and reduces the risk of the salt freezing as it is cooled in the heat exchanger.
Due to the high "redox window" of fused fluoride salts, the redox potential of the fused salt system can be changed. Fluorine-Lithium-Beryllium ("FLiBe") can be used with beryllium additions to lower the redox potential and almost eliminate corrosion. However, since beryllium is extremely toxic, special precautions must be engineered into the design to prevent its release into the environment. Many other salts can cause plumbing corrosion, especially if the reactor is hot enough to make highly reactive hydrogen.
To date, most research has focused on FLiBe, because lithium and beryllium are reasonably effective moderators and form a eutectic salt mixture with a lower melting point than each of the constituent salts. Beryllium also performs neutron doubling, improving the neutron economy. This process occurs when the beryllium nucleus emits two neutrons after absorbing a single neutron. For the fuel carrying salts, generally 1% or 2% (by mole) of UF4 is added. Thorium and plutonium fluorides have also been used.
|Material||Total neutron capture |
relative to graphite
(per unit volume)
|Moderating ratio |
(Avg. 0.1 to 10 eV)
|ZrH||~0.2||~0 if <0.14 eV, ~11449 if >0.14 eV|
Techniques for preparing and handling molten salt were first developed at ORNL.The purpose of salt purification is to eliminate oxides, sulfur and metal impurities. Oxides could result in the deposition of solid particles in reactor operation. Sulfur must be removed because of its corrosive attack on nickel-based alloys at operational temperature. Structural metal such as chromium, nickel, and iron must be removed for corrosion control.
A water content reduction purification stage using HF and helium sweep gas was specified to run at 400 °C. Oxide and sulfur contamination in the salt mixtures were removed using gas sparging of HF – H2 mixture, with the salt heated to 600 °C. (p8) Structural metal contamination in the salt mixtures were removed using hydrogen gas sparging, at 700 °C. (p26) Solid ammonium hydrofluoride was proposed as a safer alternative for oxide removal.
The possibility of online processing can be an MSR advantage. Continuous processing would reduce the inventory of fission products, control corrosion and improve neutron economy by removing fission products with high neutron absorption cross-section, especially xenon. This makes the MSR particularly suited to the neutron-poor thorium fuel cycle. Online fuel processing can introduce risks of fuel processing accidents, p15) which can trigger release of radio isotopes.(
In some thorium breeding scenarios, the intermediate product Protactinium 233
Pa would be removed from the reactor and allowed to decay into highly pure 233
U, an attractive bomb-making material. More modern designs propose to use a lower specific power or a separate thorium breeding blanket. This dilutes the protactinium to such an extent that few protactinium atoms absorb a second neutron or, via a (n, 2n) reaction (in which an incident neutron is not absorbed but instead knocks a neutron out of the nucleus), generate 232
U. Because 232
U has a short half-life and its decay chain contains hard gamma emitters, it makes the isotopic mix of uranium less attractive for bomb-making. This benefit would come with the added expense of a larger fissile inventory or a 2-fluid design with a large quantity of blanket salt.
The necessary fuel salt reprocessing technology has been demonstrated, but only at laboratory scale. A prerequisite to full-scale commercial reactor design is the R&D to engineer an economically competitive fuel salt cleaning system.
Reprocessing refers to the chemical separation of fissionable uranium and plutonium from spent fuel.Such recovery could increase the risk of nuclear proliferation. In the United States the regulatory regime has varied dramatically across administrations.
In the 1971 Molten Salt Breeder Reactor proposal, uranium reprocessing was scheduled every ten days as part of reactor operation. p181) Subsequently, a once-through fueling design was proposed that limited uranium reprocessing to every thirty years at the end of useful salt life. (p98) A mixture with 238
U was called for to make sure recovered uranium would not be weapons-grade. This design is referred to as denatured molten salt reactor. With no reprocessing, the uranium would be disposed with other fission products.
MSRs, especially those with the fuel dissolved in the salt, differ considerably from conventional reactors. Reactor core pressure can be low and the temperature much higher. In this respect an MSR is more similar to a liquid metal cooled reactor than to a conventional light water cooled reactor. MSRs are often planned as breeding reactors with a closed fuel cycle—as opposed to the once-through fuel currently used in U.S. nuclear reactors.
Safety concepts rely on a negative temperature coefficient of reactivity and a large possible temperature rise to limit reactivity excursions. As an additional method for shutdown, a separate, passively cooled container below the reactor can be included. In case of problems and for regular maintenance the fuel is drained from the reactor. This stops the nuclear reaction and acts as a second cooling system. Neutron-producing accelerators have been proposed for some super-safe subcritical experimental designs.
Cost estimates from the 1970s were slightly lower than for conventional light-water reactors.
The temperatures of some proposed designs are high enough to produce process heat for hydrogen production or other chemical reactions. Because of this, they are included in the GEN-IV roadmap for further study.
MSRs offer many potential advantages over current light water reactors:
A nuclear reactor, formerly known as an atomic pile, is a device used to initiate and control a self-sustained nuclear chain reaction. Nuclear reactors are used at nuclear power plants for electricity generation and in nuclear marine propulsion. Heat from nuclear fission is passed to a working fluid, which in turn runs through steam turbines. These either drive a ship's propellers or turn electrical generators' shafts. Nuclear generated steam in principle can be used for industrial process heat or for district heating. Some reactors are used to produce isotopes for medical and industrial use, or for production of weapons-grade plutonium. As of early 2019, the IAEA reports there are 454 nuclear power reactors and 226 nuclear research reactors in operation around the world.
The pebble-bed reactor (PBR) is a design for a graphite-moderated, gas-cooled nuclear reactor. It is a type of very-high-temperature reactor (VHTR), one of the six classes of nuclear reactors in the Generation IV initiative. The basic design of pebble-bed reactors features spherical fuel elements called pebbles. These tennis ball-sized pebbles are made of pyrolytic graphite, and they contain thousands of micro-fuel particles called TRISO particles. These TRISO fuel particles consist of a fissile material surrounded by a ceramic layer coating of silicon carbide for structural integrity and fission product containment. In the PBR, thousands of pebbles are amassed to create a reactor core, and are cooled by a gas, such as helium, nitrogen or carbon dioxide, that does not react chemically with the fuel elements. Other coolants such as FLiBe have also been suggested for implementation with pebble fuelled reactors.
A breeder reactor is a nuclear reactor that generates more fissile material than it consumes. Breeder reactors achieve this because their neutron economy is high enough to create more fissile fuel than they use, by irradiation of a fertile material, such as uranium-238 or thorium-232 that is loaded into the reactor along with fissile fuel. 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.
A fast-neutron reactor (FNR) or simply a fast reactor is a category of nuclear reactor in which the fission chain reaction is sustained by fast neutrons, as opposed to thermal neutrons used in thermal-neutron reactors. Such a reactor needs no neutron moderator, but requires fuel that is relatively rich in fissile material when compared to that required for a thermal-neutron reactor.
Passive nuclear safety is a design approach for safety features, implemented in a nuclear reactor, that does not require any active intervention on the part of the operator or electrical/electronic feedback in order to bring the reactor to a safe shutdown state, in the event of a particular type of emergency. Such design features tend to rely on the engineering of components such that their predicted behaviour would slow down, rather than accelerate the deterioration of the reactor state; they typically take advantage of natural forces or phenomena such as gravity, buoyancy, pressure differences, conduction or natural heat convection to accomplish safety functions without requiring an active power source. Many older common reactor designs use passive safety systems to a limited extent, rather, relying on active safety systems such as diesel powered motors. Some newer reactor designs feature more passive systems; the motivation being that they are highly reliable and reduce the cost associated with the installation and maintenance of systems that would otherwise require multiple trains of equipment and redundant safety class power supplies in order the achieve the same level of reliability. However, weak driving forces that power many passive safety features can pose significant challenges to effectiveness of a passive system, particularly in the short term following an accident.
Nuclear fuel is material used in nuclear power stations to produce heat to power turbines. Heat is created when nuclear fuel undergoes nuclear fission.
Generation IV reactors are a set of nuclear reactor designs currently being researched for commercial applications by the Generation IV International Forum. They are motivated by a variety of goals including improved safety, sustainability, efficiency, and cost.
The very-high-temperature reactor (VHTR), or high-temperature gas-cooled reactor (HTGR), is a Generation IV reactor concept that uses a graphite-moderated nuclear reactor with a once-through uranium fuel cycle. The VHTR is a type of high-temperature reactor (HTR) that can conceptually have an outlet temperature of 1000 °C. The reactor core can be either a "prismatic block" or a "pebble-bed" core. The high temperatures enable applications such as process heat or hydrogen production via the thermochemical sulfur–iodine cycle.
The Aircraft Nuclear Propulsion (ANP) program and the preceding Nuclear Energy for the Propulsion of Aircraft (NEPA) project worked to develop a nuclear propulsion system for aircraft. The United States Army Air Forces initiated Project NEPA on May 28, 1946. NEPA operated until May 1951, when the project was transferred to the joint Atomic Energy Commission (AEC)/USAF ANP. The USAF pursued two different systems for nuclear-powered jet engines, the Direct Air Cycle concept, which was developed by General Electric, and Indirect Air Cycle, which was assigned to Pratt & Whitney. The program was intended to develop and test the Convair X-6, but was cancelled in 1961 before that aircraft was built. The total cost of the program from 1946 to 1961 was about $1 billion.
The thorium fuel cycle is a nuclear fuel cycle that uses an isotope of thorium, 232
, as the fertile material. In the reactor, 232
is transmuted into the fissile artificial uranium isotope 233
which is the nuclear fuel. Unlike natural uranium, natural thorium contains only trace amounts of fissile material, which are insufficient to initiate a nuclear chain reaction. Additional fissile material or another neutron source is necessary to initiate the fuel cycle. In a thorium-fuelled reactor, 232
absorbs neutrons to produce 233
. This parallels the process in uranium breeder reactors whereby fertile 238
absorbs neutrons to form fissile 239
. Depending on the design of the reactor and fuel cycle, the generated 233
either fissions in situ or is chemically separated from the used nuclear fuel and formed into new nuclear fuel.
Alvin Martin Weinberg was an American nuclear physicist who was the administrator at Oak Ridge National Laboratory (ORNL) during and after the Manhattan Project. He came to Oak Ridge, Tennessee, in 1945 and remained there until his death in 2006. He was the first to use the term "Faustian bargain" to describe nuclear energy.
The Molten-Salt Reactor Experiment (MSRE) was an experimental molten salt reactor at the Oak Ridge National Laboratory (ORNL) researching this technology through the 1960s; constructed by 1964, it went critical in 1965 and was operated until 1969.
The liquid fluoride thorium reactor is a type of molten salt reactor. LFTRs use the thorium fuel cycle with a fluoride-based, molten, liquid salt for fuel. In a typical design, the liquid is pumped between a critical core and an external heat exchanger where the heat is transferred to a nonradioactive secondary salt. The secondary salt then transfers its heat to a steam turbine or closed-cycle gas turbine.
The FUJI molten salt reactor is a proposed molten-salt-fueled thorium fuel cycle thermal breeder reactor, using technology similar to the Oak Ridge National Laboratory's Molten Salt Reactor Experiment – liquid fluoride thorium reactor. It was being developed by the Japanese company International Thorium Energy & Molten-Salt Technology (IThEMS), together with partners from the Czech Republic. As a breeder reactor, it converts thorium into the nuclear fuel uranium-233. To achieve reasonable neutron economy, the chosen single-salt design results in significantly larger feasible size than a two-salt reactor. Like all molten salt reactors, its core is chemically inert and under low pressure, helping to prevent explosions and toxic releases. The proposed design is rated at 200 MWe output. The IThEMS consortium planned to first build a much smaller MiniFUJI 10 MWe reactor of the same design once it had secured an additional $300 million in funding.
FLiBe is a molten salt made from a mixture of lithium fluoride (LiF) and beryllium fluoride (BeF2). It is both a nuclear reactor coolant and solvent for fertile or fissile material. It served both purposes in the Molten-Salt Reactor Experiment (MSRE) at the Oak Ridge National Laboratory.
Thorium-based nuclear power generation is fueled primarily by the nuclear fission of the isotope uranium-233 produced from the fertile element thorium. According to proponents, a thorium fuel cycle offers several potential advantages over a uranium fuel cycle—including much greater abundance of thorium found on Earth, superior physical and nuclear fuel properties, and reduced nuclear waste production. However, development of thorium power has significant start-up costs. Proponents also cite the low weaponization potential as an advantage of thorium due to how difficult it is to weaponize the specific uranium-233/232 and plutonium-238 isotopes produced by thorium reactors, while critics say that development of breeder reactors in general increases proliferation concerns. As of 2020, there are no operational thorium reactors in the world.
The Integral Molten Salt Reactor (IMSR) is designed for the small modular reactor (SMR) market. It employs molten salt reactor technology which is being developed by the Canadian company Terrestrial Energy. It is based closely on the denatured molten salt reactor (DMSR), a reactor design from Oak Ridge National Laboratory. It also incorporates elements found in the SmAHTR, a later design from the same laboratory. The IMSR belongs to the DMSR class of molten salt reactors (MSR) and hence is a "burner" reactor that employs a liquid fuel rather than a conventional solid fuel; this liquid contains the nuclear fuel and also serves as primary coolant.
Transatomic Power was an American company that designed Generation IV nuclear reactors based on molten salt reactor (MSR) technology.
The stable salt reactor (SSR) is a nuclear reactor design under development by Moltex Energy Ltd, based in the United Kingdom and Canada.
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