|RBMK Reactor Class|
View of the Smolensk Nuclear Power Plant site, with three operational RBMK-1000 reactors. A fourth reactor was cancelled before completion.
|Generation||Generation II reactor|
|Reactor concept||Graphite-moderated boiling water reactor|
|Reactor line||RBMK (Reaktor Bolshoy Moshchnosti Kanalniy)|
|Status|| 26 blocks: |
|Main parameters of the reactor core|
|Fuel (fissile material)||235U (NU/SEU/LEU)|
|Neutron energy spectrum||Thermal|
|Primary control method||Control rods|
|Primary coolant||Liquid (light water)|
|Primary use||Generation of electricity and production of weapon grade plutonium|
|Power (thermal)||RBMK-1000: 3,200 MWth|
RBMK-1500: 4,800 MWth
RBMKP-2400: 6,500 MWth
|Power (electric)||RBMK-1000: 1,000 MWe|
RBMK-1500: 1,500 MWe
RBMKP-2400: 2,400 MWe
The RBMK (Russian : Реактор Большой Мощности Канальный, РБМК; Reaktor Bolshoy Moshchnosti Kanalnyy, "High Power Channel-type Reactor") is a class of graphite-moderated nuclear power reactor designed and built by the Soviet Union. The name refers to its unusual design where, instead of a large steel pressure vessel surrounding the entire core, each fuel assembly is enclosed in an individual 8 cm diameter pipe (called a "channel") which allows the flow of cooling water around the fuel.
Russian is an East Slavic language, which is an official language in the Russian Federation, Belarus, Kazakhstan and Kyrgyzstan, as well as being widely used throughout Eastern Europe, the Baltic states, the Caucasus and Central Asia. It was the de facto language of the Soviet Union until its dissolution on 25 December 1991. Although nearly three decades have passed since the breakup of the Soviet Union, Russian is used in official capacity or in public life in all the post-Soviet nation-states, as well as in Israel and Mongolia.
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 Soviet Union, officially known as the Union of Soviet Socialist Republics (USSR), was a federal sovereign state in northern Eurasia that existed from 1922 to 1991. Nominally a union of multiple national Soviet republics, in practice its government and economy were highly centralized. The country was a one-party state, governed by the Communist Party with Moscow as its capital in its largest republic, the Russian Soviet Federative Socialist Republic. Other major urban centers were Leningrad, Kiev, Minsk, Tashkent, Alma-Ata, and Novosibirsk. It spanned over 10,000 kilometers (6,200 mi) east to west across 11 time zones, and over 7,200 kilometers (4,500 mi) north to south. Its territory included much of Eastern Europe, as well as part of Northern Europe and all of Northern and Central Asia. It had five climate zones: tundra, taiga, steppes, desert and mountains.
The RBMK is an early Generation II reactor and the oldest commercial reactor design still in wide operation. Certain aspects of the RBMK reactor design, such as the active removal of decay heat, the positive void coefficient properties, the 4.5 m (14 ft 9 in) graphite displacer ends of the control rods and instability at low power levels, contributed to the 1986 Chernobyl disaster, in which an RBMK experienced a very large reactivity excursion, leading to a steam and hydrogen explosion, a large fire and subsequent meltdown. Radioactivity was released over a large portion of Europe. The disaster prompted worldwide calls for the reactors to be completely decommissioned; however, there is still considerable reliance on RBMK facilities for power in Russia. Most of the flaws in the design of RBMK-1000 reactors were corrected after the Chernobyl accident and a dozen reactors have since been operating without any serious incidents for over twenty years. While nine RBMK blocks under construction were cancelled after the Chernobyl disaster, and the last of three remaining RBMK blocks at the Chernobyl Nuclear Power Plant was shut down in 2000, as of 2019 there were still 10 RBMK reactors and three small EGP-6 graphite moderated light-water reactors operating in Russia, though all have been retrofitted with a number of safety updates. Only two RBMK blocks were started after 1986: Ignalina-2 and Smolensk-3.
A generation II reactor is a design classification for a nuclear reactor, and refers to the class of commercial reactors built until the end of the 1990s. Prototypical generation II reactors include the PWR, CANDU, BWR, AGR, older VVER and RBMK.
Decay heat is the heat released as a result of radioactive decay. This heat is produced as an effect of radiation on materials: the energy of the alpha, beta or gamma radiation is converted into the thermal movement of atoms.
In nuclear engineering, the void coefficient is a number that can be used to estimate how much the reactivity of a nuclear reactor changes as voids form in the reactor moderator or coolant. Reactivity, in the nuclear engineering sense, measures the degree of change in neutron multiplication in a reactor core. Reactivity is directly related to the tendency of the reactor core to change power level: if reactivity is positive, the core power tends to increase; if it is negative, the core power tends to decrease; if it is zero, the core power tends to remain stable. The reactivity of the core may be adjusted by the reactor control system in order to obtain a desired power level change. It can be compared to the reaction of an automobile as conditions around it change, and therefore the corresponding counter-measure that the driver applies to maintain road speed or execute a desired maneuver.
The only differences between RBMK-1000 and RBMK-1500 reactors are that the RBMK-1500 is cooled with less water (thus more of the water turns into steam), and it uses less uranium. The only reactors of this type and power output are the ones at Ignalina Nuclear Power Plant. The RBMKP-2400 is rectangular instead of cylindrical, and it was intended to be made in sections at a factory for assembly in situ. It was designed to have a power output of 2400 MWe. No reactor with this power output has ever been built, with the most powerful one currently being as of 2018 the 1750 MWe EPR.
The Ignalina Nuclear Power Plant is a decommissioned two-unit RBMK-1500 nuclear power station in Visaginas Municipality, Lithuania. It was named after the nearby city of Ignalina. Due to the plant's similarities to the infamous Chernobyl Nuclear Power Plant in both reactor design and lack of a robust containment building, Lithuania agreed to close the plant as part of its accession agreement to the European Union. Unit 1 was closed in December 2004; Unit 2, which counted for 25% of Lithuania's electricity generating capacity and supplied about 70% of Lithuania's electrical demand, was closed on December 31, 2009. Proposals have been made to construct a new nuclear power plant at the same site. However, plans have not materialised since then, and the country is one of the most active supporters of renewable energy.
In situ is a Latin phrase that translates literally to "on site" or "in position." It can mean "locally", "on site", "on the premises", or "in place" to describe where an event takes place and is used in many different contexts. For example, in fields such as physics, geology, chemistry, or biology, in situ may describe the way a measurement is taken, that is, in the same place the phenomenon is occurring without isolating it from other systems or altering the original conditions of the test.
The EPR is a third generation pressurised water reactor design. It has been designed and developed mainly by Framatome and Électricité de France (EDF) in France, and Siemens in Germany. In Europe this reactor design was called European Pressurised Reactor, and the internationalised name was Evolutionary Power Reactor, but it is now simply named EPR.
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The RBMK was the culmination of the Soviet nuclear power program to produce a water-cooled power reactor with dual-use potential based on their graphite-moderated plutonium production military reactors.
Plutonium is a radioactive chemical element with the symbol Pu and atomic number 94. It is an actinide metal of silvery-gray appearance that tarnishes when exposed to air, and forms a dull coating when oxidized. The element normally exhibits six allotropes and four oxidation states. It reacts with carbon, halogens, nitrogen, silicon, and hydrogen. When exposed to moist air, it forms oxides and hydrides that can expand the sample up to 70% in volume, which in turn flake off as a powder that is pyrophoric. It is radioactive and can accumulate in bones, which makes the handling of plutonium dangerous.
The first of these, Obninsk AM-1 ("Атом Мирный", Atom Mirny, Russian for "Atoms for Peace") generated 5 MW of electricity from 30 MW thermal power, and supplied Obninsk from 1954 until 1959. Subsequent prototypes were the AMB-100 reactor and AMB-200 reactor both at Beloyarsk Nuclear Power Station.
Obninsk Nuclear Power Plant was built in the "Science City" of Obninsk, Kaluga Oblast, about 110 km (68 mi) southwest of Moscow, Soviet Union. Connected to the power grid in June 1954, Obninsk was the first grid-connected nuclear power plant in the world, i.e. the first nuclear reactor that produced electricity industrially, albeit at small scale. It was located at the Institute of Physics and Power Engineering. The plant is also known as APS-1 Obninsk. It remained in operation between 1954 and 2002, although its production of electricity for the grid ceased in 1959; thereafter it functioned as a research and isotope production plant only.
"Atoms for Peace" was the title of a speech delivered by U.S. President Dwight D. Eisenhower to the UN General Assembly in New York City on December 8, 1953.
I feel impelled to speak today in a language that in a sense is new – one which I, who have spent so much of my life in the military profession, would have preferred never to use.
That new language is the language of atomic warfare.
The Beloyarsk Nuclear Power Station was the third of the Soviet Union's nuclear plants. It is situated by Zarechny in Sverdlovsk Oblast, Russia. Zarechny township was created to service the station, which is named after the Beloyarsky District. The closest city is Yekaterinburg.
By using a minimalist design that used regular (light) water for cooling and graphite for moderation, it was possible to use natural uranium for fuel (instead of the considerably more expensive enriched uranium). This allowed for an extraordinarily large and powerful reactor that was also cheap enough to be built in large numbers and simple enough to be maintained and operated by local personnel. For example, the RBMK reactors at the Ignalina Nuclear Power Plant in Lithuania were rated at 1500 MWe each, a very large size for the time and even for the early 21st century.
Water is a transparent, tasteless, odorless, and nearly colorless chemical substance, which is the main constituent of Earth's hydrosphere, and the fluids of most living organisms. It is vital for all known forms of life, even though it provides no calories or organic nutrients. Its chemical formula is H2O, meaning that each of its molecules contains one oxygen and two hydrogen atoms, connected by covalent bonds. Water is the name of the liquid state of H2O at standard ambient temperature and pressure. It forms precipitation in the form of rain and aerosols in the form of fog. Clouds are formed from suspended droplets of water and ice, its solid state. When finely divided, crystalline ice may precipitate in the form of snow. The gaseous state of water is steam or water vapor. Water moves continually through the water cycle of evaporation, transpiration (evapotranspiration), condensation, precipitation, and runoff, usually reaching the sea.
In nuclear engineering, a neutron moderator is a medium that reduces the speed of fast neutrons, ideally without capturing any, leaving them as thermal neutrons with only minimal (thermal) kinetic energy. These thermal neutrons are immensely more suscepible than fast neutrons to propagate a nuclear chain reaction of uranium-235 or other fissile isotope by colliding with their atomic nucleus.
Natural uranium refers to uranium with the same isotopic ratio as found in nature. It contains 0.711% uranium-235, 99.284% uranium-238, and a trace of uranium-234 by weight (0.0055%). Approximately 2.2% of its radioactivity comes from uranium-235, 48.6% from uranium-238, and 49.2% from uranium-234.
A 45-year lifetime is envisaged for many of the units, after mid-life refurbishment.
The reactor pit is made of reinforced concrete and has dimensions 21.6 by 21.6 by 25.5 metres (71 ft × 71 ft × 84 ft). It houses the vessel of the reactor, made of a cylindrical wall and top and bottom metal plates. The vessel contains the graphite stack and is filled with a helium-nitrogen mixture for providing an inert atmosphere for the graphite and for mediation of heat transfer from the graphite to the coolant channels. The moderator blocks are made of nuclear graphite the dimensions of which are 25 by 25 centimetres (9.8 in × 9.8 in) on the plane perpendicular to the channels and with several longitudinal dimensions of between 20 centimetres (7.9 in) and 60 centimetres (24 in) depending on the location in the stack. There are holes of 11.4 cm (4.5 in) diameter through the longitudinal axis of the blocks for the fuel and control channels. The blocks are stacked inside the reactor vessel into a cylindrical core with a diameter and height of 14 by 8 metres (46 ft × 26 ft). The maximum allowed temperature of the graphite is up to 730 °C (1,350 °F).
The reactor vessel is a steel cylinder with a diameter and height of 14.52 by 9.75 metres (47.6 ft × 32.0 ft), and a wall thickness of 16 mm (0.63 in). In order to absorb axial thermal expansion loads, it is equipped with a bellows compensator.
The moderator is surrounded by a cylindrical water tank, a welded structure with 3 cm (1.2 in) thick walls, an inner diameter of 16.6 m (54 ft 6 in) and an outer diameter of 19 m (62 ft 4 in), internally divided to 16 vertical compartments. The water is supplied to the compartments from the bottom and removed from the top; the water can be used for emergency reactor cooling. The tank contains thermocouples for sensing the water temperature and ion chambers for monitoring the reactor power. The tank, sand layer, and concrete of the reactor pit serve as additional biological shields.
The top of the reactor is covered by the upper biological shield (UBS), also called "Schema E", Pyatachok, or, after the explosion (of Chernobyl Reactor 4), Elena. The UBS is a cylindrical disc of 3 m × 17 m (9.8 ft × 55.8 ft) in size. It is penetrated by standpipes for fuel and control channel assemblies. The top and bottom are covered with 4 cm (1.57 in) thick steel plates, welded to be helium-tight, and additionally joined by structural supports. The space between the plates and pipes is filled with serpentinite, a rock containing significant amounts of bound water. The disk is supported on 16 rollers, located on the upper side of the reinforced cylindrical water tank. The structure of the UBS supports the fuel and control channels, the floor above the reactor in the central hall, and the steam-water pipes.
Below the bottom of the reactor core there is the lower biological shield (LBS), similar to the UBS, but only 2 m × 14.5 m (6.6 ft × 47.6 ft) in size. It is penetrated by the tubes for the lower ends of the pressure channels and carries the weight of the graphite stack and the coolant inlet piping. A steel structure, two heavy plates intersecting in right angle under the center of the LBS and welded to the LBS, supports the LBS and transfers the mechanical load to the building.
Above the UBS, there is the upper shield cover; its top surface is the floor of the central hall. It serves as part of the biological shield and for thermal insulation of the reactor space. Its center area above the reactor channel consists of individual removable steel-graphite plugs, located over the tops of the channels.
The fuel channels consist of welded zircaloy pressure tubes 8 cm (3.1 in) in inner diameter with 4 mm (0.16 in) thick walls, led through the channels in the center of the graphite moderator blocks. The top and bottom parts of the tubes are made of stainless steel, and joined with the central zircaloy segment with zirconium-steel alloy couplings. The pressure tube is held in the graphite stack channels with two alternating types of 20 mm (0.79 in) high split graphite rings; one is in direct contact with the tube and has 1.5 mm (0.059 in) clearance to the graphite stack, the other one is directly touching the graphite stack and has 1.3 mm (0.051 in) clearance to the tube; this assembly reduces transfer of mechanical loads caused by neutron-induced swelling, thermal expansion of the blocks, and other factors to the pressure tube, while facilitating heat transfer from the graphite blocks. The pressure tubes are welded to the top and bottom plates of the reactor vessel.
While most of the heat energy from the fission process is generated in the fuel rods, approximately 5.5% is deposited in the graphite blocks as they moderate the fast neutrons formed from fission. This energy must be removed to avoid overheating the graphite. About 80–85% of the energy deposited in the graphite is removed by the fuel rod coolant channels, using conduction via the graphite rings. The rest of the graphite heat is removed from the control rod channels by forced gas circulation.
There are 1693 fuel channels and 170 control rod channels in the first generation RBMK reactor cores. Second generation reactor cores (such as Chernobyl-4) have 1661 fuel channels and 211 control rod channels.
The fuel assembly is suspended in the fuel channel on a bracket, with a seal plug. The seal plug has a simple design, to facilitate its removal and installation by the remotely controlled refueling machine.
The fuel channels may, instead of fuel, contain fixed neutron absorbers, or be filled completely with cooling water.
The small clearance between the pressure channel and the graphite block makes the graphite core susceptible to damage. If a pressure channel deforms, e.g. by too high an internal pressure, the deformation can cause significant pressure loads on the graphite blocks and lead to damage.
The fuel pellets are made of uranium dioxide powder, sintered with a suitable binder into barrels 11.5 mm (0.45 in) in diameter and 15 mm (0.59 in) long. The material may contain added europium oxide as a burnable nuclear poison to lower the reactivity differences between a new and partially spent fuel assembly. To reduce thermal expansion issues and interaction with the cladding, the pellets have hemispherical indentations. A 2 mm (0.079 in) hole through the axis of the pellet serves to reduce the temperature in the center of the pellet and facilitates removal of gaseous fission products. The enrichment level is 2% (0.4% for the end pellets of the assemblies). Maximum allowable temperature of the fuel pellet is 2,100 °C (3,810 °F).
The fuel rods are zircaloy (1% niobium) tubes 13.6 mm (0.54 in) in outer diameter, 0.825 mm (0.0325 in) thick. The rods are filled with helium at 0.5 MPa and hermetically sealed. Retaining rings help to seat the pellets in the center of the tube and facilitate heat transfer from the pellet to the tube. The pellets are axially held in place by a spring. Each rod contains 3.5 kg (7.7 lb) of fuel pellets. The fuel rods are 3.64 m (11 ft 11 in) long, with 3.4 m (11 ft 2 in) of that being the active length. The maximum allowed temperature of a fuel rod is 600 °C (1,112 °F).
The fuel assemblies consist of two sets ("sub-assemblies") with 18 fuel rods and 1 carrier rod. The fuel rods are arranged along the central carrier rod, which has an outer diameter of 1.3 cm (0.5 in). All rods of a fuel assembly are held in place with 10 stainless steel spacers separated by 360 mm (14.2 in) distance. The two sub-assemblies are joined with a cylinder at the center of the assembly; during the operation of the reactor, this dead space without fuel lowers the neutron flux in the central plane of the reactor. The total mass of uranium in the fuel assembly is 114.7 kg (253 lb). The fuel burnup is 20 MW·d/kg. The total length of the fuel assembly is 10.025 m (32 ft 10.7 in), with 6.862 m (22 ft 6.2 in) of the active region.
In addition to the regular fuel assemblies, there are instrumented ones, containing neutron flux detectors in the central carrier. In this case, the rod is replaced with a tube with wall thickness of 2.5 mm (0.098 in); and outer diameter of 15 mm (0.6 in).
Unlike the rectangular PWR/BWR fuel assemblies, the RBMK fuel assembly is cylindrical to fit the round pressure channels.
The refueling machine is mounted on a gantry crane and remotely controlled. The fuel assemblies can be replaced without shutting down the reactor, a factor significant for production of weapon-grade plutonium and, in a civilian context, for better reactor uptime. When a fuel assembly has to be replaced, the machine is positioned above the fuel channel: then it mates to the latter, equalizes pressure within, pulls the rod, and inserts a fresh one. The spent rod is then placed in a cooling pond. The capacity of the refueling machine with the reactor at nominal power level is two fuel assemblies per day, with peak capacity of five per day.
The total amount of fuel under stationary conditions is 192 tons.
Most of the reactor control rods are inserted from above; 24 shortened rods are inserted from below and are used to augment the axial power distribution control of the core. With the exception of 12 automatic rods, the control rods have a 4.5 m (14 ft 9 in) long graphite section at the end, separated by a 1.25 m (4 ft 1 in) long telescope (which creates a water-filled space between the graphite and the absorber), and a boron carbide neutron absorber section. The role of the graphite section, known as "displacer", is to enhance the difference between the neutron flux attenuation levels of inserted and retracted rods, as the graphite displaces water that would otherwise act as a neutron absorber, although much weaker than boron carbide; a control rod channel filled with graphite absorbs fewer neutrons than when filled with water, so the difference between inserted and retracted control rod is increased. When the control rod is fully retracted, the graphite displacer is located in the middle of the core height, with 1.25 m of water at each of its ends. The displacement of water in the lower 1.25 m of the core as the rod moves down causes a local increase of reactivity in the bottom of the core as the graphite part of the control rod passes that section. This "positive scram" effect was discovered in 1983 at the Ignalina Nuclear Power Plant. The control rod channels are cooled by an independent water circuit and kept at 40–70 °C (104–158 °F). The narrow space between the rod and its channel hinders water flow around the rods during their movement and acts as a fluid damper, which is the primary cause of their slow insertion time (nominally 18–21 seconds for the RCPS rods, or about 0.4 m/s). After the Chernobyl disaster, the control rod servos on other RBMK reactors were exchanged to allow faster rod movements, and even faster movement was achieved by cooling of the control rod tubes by a thin layer of water while letting the rods themselves move in gas.
The division of the control rods between manual and emergency protection groups was arbitrary; the rods could be reassigned from one system to another during reactor operation without technical or organizational problems.
Additional static boron-based absorbers are inserted into the core when it is loaded with fresh fuel. About 240 absorbers are added during initial core loading. These absorbers are gradually removed with increasing burnup. The reactor's void coefficient depends on the core content; it ranges from negative with all the initial absorbers to positive when they are all removed.
The normal reactivity margin is 43–48 control rods.
The reactor operates in a helium–nitrogen atmosphere (70–90% He, 10–30% N2).The gas circuit is composed of a compressor, aerosol and iodine filters, adsorber for carbon dioxide, carbon monoxide, and ammonia, a holding tank for allowing the gaseous radioactive products to decay before being discharged, an aerosol filter to remove solid decay products, and a ventilator stack, the iconic chimney above the plant building. The gas is injected to the stack from the bottom in a low flow rate, and exits from the standpipe of each channel via an individual pipe. The moisture and temperature of the outlet gas is monitored; an increase of them is an indicator of a coolant leak.
The reactor has two independent cooling circuits, each having four main circulating pumps (three operating, one standby). The cooling water is fed to the reactor through lower water lines to a common pressure header (one for each cooling circuit), which is split to 22 group distribution headers, each feeding 38–41 pressure channels through the core, where the feedwater boils. The mixture of steam and water is led by the upper steam lines, one for each pressure channel, from the reactor top to the steam separators, pairs of thick horizontal drums located in side compartments above the reactor top; each has 2.8 m (9 ft 2 in) diameter, 31 m (101 ft 8 in) length, wall thickness of 10 cm (3.9 in), and weighs 240 t (260 short ton s). Steam, with steam quality of about 15%, is taken from the top of the separators by two steam collectors per separator, combined, and led to two turbogenerators in the turbine hall, then to condensers, reheated to 165 °C (329 °F), and pumped by the condensate pumps to deaerators, where remains of gaseous phase and corrosion-inducing gases are removed. The resulting feedwater is led to the steam separators by feedwater pumps and mixed with water from them at their outlets. From the bottom of the steam separators, the feedwater is led by 12 downpipes (from each separator) to the suction headers of the main circulation pumps, and back into the reactor. There is an ion exchange system included in the loop to remove impurities from the feedwater.
The turbine consists of one high-pressure rotor and four low-pressure ones. Five low-pressure separators-preheaters are used to heat steam with fresh steam before being fed to the next stage of the turbine. The uncondensed steam is fed into a condenser, mixed with condensate from the separators, fed by the first-stage condensate pump to a chemical purifier, then by a second-stage condensate pump to four deaerators where dissolved and entrained gases are removed; deaerators also serve as storage tanks for feedwater. From the deaerators, the water is pumped through filters and into the bottom parts of the steam separator drums.
The main circulating pumps have the capacity of 5,500–12,000 m3/h and are powered by 6 kV electric motors. The normal coolant flow is 8000 m3/h per pump; this is throttled down by control valves to 6000–7000 m3/h when the reactor power is below 500 MWt. Each pump has a flow control valve and a backflow preventing check valve on the outlet, and shutoff valves on both inlet and outlet. Each of the pressure channels in the core has its own flow control valve so that the temperature distribution in the reactor core can be optimized. Each channel has a ball type flow meter.
The nominal coolant flow through the reactor is 46,000–48,000 m3/h. The steam flow at full power is 5,440–5,600 t (6,000–6,170 short tons)/h.
The nominal temperature of the cooling water at the inlet of the reactor is about 265–270 °C (509–518 °F) and the outlet temperature 284 °C (543 °F), at pressure in the drum separator of 6.9 megapascals (69 bar; 1,000 psi). The pressure and the inlet temperature determine the height at which the boiling begins in the reactor; if the coolant temperature is not sufficiently below its boiling point at the system pressure, the boiling starts at the very bottom part of the reactor instead of its higher parts. With few absorbers in the reactor core, such as during the Chernobyl accident, the positive void coefficient of the reactor makes the reactor very sensitive to the feedwater temperature. Bubbles of boiling water lead to increased power, which in turn increases the formation of bubbles. After 1986 absorbers were introduced in the fuel assembly, permanently assuring a negative void coefficient at the cost of higher enrichment requirements of the uranium fuel.
If the coolant temperature is too close to its boiling point, cavitation can occur in the pumps and their operation can become erratic or even stop entirely. The feedwater temperature is dependent on the steam production; the steam phase portion is led to the turbines and condensers and returns significantly cooler (155–165 °C (311–329 °F)) than the water returning directly from the steam separator (284 °C). At low reactor power, therefore, the inlet temperature may become dangerously high. The water is kept below the saturation temperature to prevent film boiling and the associated drop in heat transfer rate.
The reactor is tripped in cases of high or low water level in the steam separators (with two selectable low-level thresholds); high steam pressure; low feedwater flow; loss of two main coolant pumps on either side. These trips can be manually disabled.
The level of water in the steam separators, the percentage of steam in the reactor pressure tubes, the level at which the water begins to boil in the reactor core, the neutron flux and power distribution in the reactor, and the feedwater flow through the core have to be carefully controlled. The level of water in the steam separator is mainly controlled by the feedwater supply, with the deaerator tanks serving as a water reservoir.
The maximum allowed heat-up rate of the reactor and the coolant is 10 °C (18 °F)/h; the maximum cool-down rate is 30 °C (54 °F)/h.
The reactor is equipped with an emergency core cooling system (ECCS), consisting of dedicated water reserve tank, hydraulic accumulators, and pumps. ECCS piping is integrated with the normal reactor cooling system. In case of total loss of power, the ECCS pumps are supposed to be powered by the rotational momentum of the turbogenerator rotor for the time before the diesel generators come online. The Chernobyl disaster occurred during a botched test of this system. The ECCS has three systems, connected to the coolant system headers. In case of damage, the first ECCS subsystem provides cooling for up to 100 seconds to the damaged half of the coolant circuit (the other half is cooled by the main circulation pumps), and the other two subsystems then handle long-term cooling of the reactor.
The short-term ECCS subsystem consists of two groups of six accumulator tanks, containing water blanketed with nitrogen under pressure of 10 megapascals (1,500 psi), connected by fast-acting valves to the reactor. Each group can supply 50% of the maximum coolant flow to the damaged half of the reactor. The third group is a set of electrical pumps drawing water from the deaerators. The short-term pumps can be powered by the spindown of the main turbogenerators.
ECCS for long-term cooling of the damaged circuit consists of three pairs of electrical pumps, drawing water from the pressure suppression pools; the water is cooled by the plant service water by means of heat exchangers in the suction lines. Each pair is able to supply half of the maximum coolant flow. ECCS for long-term cooling of the intact circuit consists of three separate pumps drawing water from the condensate storage tanks, each able to supply half of the maximum flow. The ECCS pumps are powered from the essential internal 6 kV lines, backed up by diesel generators. Some valves that require uninterrupted power are also backed up by batteries.
The distribution of power density in the reactor is measured by ionization chambers located inside and outside the core. The physical power density distribution control system (PPDDCS) has sensors inside the core; the reactor control and protection system (RCPS) uses sensors in the core and in the lateral biological shield tank. The external sensors in the tank are located around the reactor middle plane, therefore do not indicate axial power distribution nor information about the power in the central part of the core. There are over 100 radial and 12 axial power distribution monitors, employing self-powered detectors. Reactivity meters and removable startup chambers are used for monitoring of reactor startup. Total reactor power is recorded as the sum of the currents of the lateral ionization chambers. The moisture and temperature of the gas circulating in the channels is monitored by the pressure tube integrity monitoring system.
The PPDDCS and RCPS are supposed to complement each other. The RCPS system consists of 211 movable control rods. Both systems, however, have deficiencies, most noticeably at low reactor power levels. The PPDDCS is designed to maintain reactor power density distribution between 10 and 120% of nominal levels and to control the total reactor power between 5 and 120% of nominal levels. The LAC-LAP (local automatic control and local automatic protection) RPCS subsystems rely on ionization chambers inside the reactor and are active at power levels above 10%. Below those levels, the automatic systems are disabled and the in-core sensors are not accessible. Without the automatic systems and relying only on the lateral ionization chambers, control of the reactor becomes very difficult; the operators do not have sufficient data to control the reactor reliably and have to rely on their intuition. During startup of a reactor with a poison-free core this lack of information can be manageable because the reactor behaves predictably, but a non-uniformly poisoned core can cause large nonhomogenities of power distribution, with potentially catastrophic results.
The reactor emergency protection system (EPS) was designed to shut down the reactor when its operational parameters are exceeded. The design accounted for steam collapse in the core when the fuel element temperature falls below 265 °C, coolant vaporization in fuel channels in cold reactor state, and sticking of some emergency protection rods. However, the slow insertion speed of the control rods, together with their design causing localized positive reactivity as the displacer moves through the lower part of the core, created a number of possible situations where initiation of the EPS could itself cause or aggravate a reactor runaway.
The computer system for calculation of the reactivity margin was collecting data from about 4,000 sources. Its purpose was to assist the operator with steady-state control of the reactor. Ten to fifteen minutes were required to cycle through all the measurements and calculate the results.[ citation needed ]
The operators could disable some safety systems, reset or suppress some alarm signals, and bypass automatic scram, by attaching patch cables to accessible terminals. This practice was allowed under some circumstances.
The reactor is equipped with a fuel rod leak detector. A scintillation counter detector, sensitive to energies of short-lived fission products, is mounted on a special dolly and moved over the outlets of the fuel channels, issuing an alert if increased radioactivity is detected in the steam-water flow.
The RBMK design was built primarily to be powerful, quick to build and easy to maintain. Full physical containment structures for each reactor would have more than doubled the cost and construction time of each plant, and since the design had been certified by the Soviet nuclear science ministry as inherently safe when operated within established parameters, the Soviet authorities assumed proper adherence to doctrine by workers would make any accident impossible. Additionally, RBMK reactors were designed to allow fuel rods to be changed at full power without shutting down (as in the pressurized heavy water CANDU reactor), both for refueling and for plutonium production (for nuclear weapons). This required large cranes above the core. As the RBMK reactor is very tall (about 7 m (23 ft 0 in)), the cost and difficulty of building a heavy containment structure prevented the building of additional emergency containment structures for pipes on top of the reactor. In the Chernobyl accident, the pressure rose to levels high enough to blow the top off the reactor, breaking open the fuel channels in the process and starting a massive fire when air contacted the superheated graphite core. After the Chernobyl accident, some RBMK reactors were retrofitted with a partial containment structure (in lieu of a full containment building), which surround the fuel channels with water jackets in order to capture any radioactive particles released.
The bottom part of the reactor is enclosed in a watertight compartment. There is a space between the reactor bottom and the floor. The reactor cavity overpressure protection system consists of steam relief assemblies embedded in the floor and leading to Steam Distributor Headers covered with rupture discs and opening into the Steam Distribution Corridor below the reactor, on level +6. The floor of the corridor contains entrances of a large number of vertical pipes, leading to the bottoms of the Pressure Suppression Pools ("bubbler" pools) located on levels +3 and +0. In the event of an accident, which was predicted to be at most a rupture of one or two pressure channels, the steam was to be bubbled through the water and condensed there, reducing the overpressure in the leaktight compartment. The flow capacity of the pipes to the pools limited the protection capacity to simultaneous rupture of two pressure channels; a higher number of failures would cause pressure buildup sufficient to lift the cover plate ("Structure E", after the explosion nicknamed "Elena"), sever the rest of the fuel channels, destroy the control rod insertion system, and potentially also withdraw control rods from the core. MPa. The distribution headers and inlets enclosures can handle 0.08 MPa and are vented via check valves to the leaktight compartment. The reactor cavity can handle overpressure of 0.18 MPa and is vented via check valves to the leaktight compartment. The pressure suppression system can handle a failure of one reactor channel, a pump pressure header, or a distribution header. Leaks in the steam piping and separators are not handled, except for maintaining slightly lower pressure in the riser pipe gallery and the steam drum compartment than in the reactor hall. These spaces are also not designed to withstand overpressure. The steam distribution corridor contains surface condensers. The fire sprinkler systems, operating during both accident and normal operation, are fed from the pressure suppression pools through heat exchangers cooled by the plant service water, and cool the air above the pools. Jet coolers are located in the topmost parts of the compartments; their role is to cool the air and remove the steam and radioactive aerosol particles.The containment was designed to handle failures of the downcomers, pumps, and distribution and inlet of the feedwater. The leaktight compartments around the pumps can withstand overpressure of 0.45
Hydrogen removal from the leaktight compartment is performed by removal of 800 m3/h of air, its filtration, and discharge into the atmosphere. The air removal is stopped automatically in case of a coolant leak and has to be reinstated manually. Hydrogen is present during normal operation due to leaks of coolant (assumed to be up to 2 t (2.2 short tons) per hour).
For the nuclear systems described here, the Chernobyl Nuclear Power Plant is used as the example.
The power plant is connected to the 330 kV and 750 kV electrical grid. The block has two electrical generators connected to the 750 kV grid by a single generator transformer. The generators are connected to their common transformer by two switches in series. Between them, the unit transformers are connected to supply power to the power plant's own systems; each generator can therefore be connected to the unit transformer to power the plant, or to the unit transformer and the generator transformer to also feed power to the grid. The 330 kV line is normally not used, and serves as an external power supply, connected by a station transformer to the power plant's electrical systems. The plant can be powered by its own generators, or get power from the 750 kV grid through the generator transformer, or from the 330 kV grid via the station transformer, or from the other power plant block via two reserve busbars. In case of total external power loss, the essential systems can be powered by diesel generators. Each unit transformer is connected to two 6 kV main power boards, A and B (e.g. 7A, 7B, 8A, 8B for generators 7 and 8), powering principal non-essential drivers and connected to transformers for the 4 kV main power and the 4 kV reserve busbar. The 7A, 7B, and 8B boards are also connected to the three essential power lines (namely for the coolant pumps), each also having its own diesel generator. In case of a coolant circuit failure with simultaneous loss of external power, the essential power can be supplied by the spinning down turbogenerators for about 45–50 seconds, during which time the diesel generators should start up. The generators are started automatically within 15 seconds at loss of off-site power.
The electrical energy is generated by a pair of 500 MW hydrogen-cooled turbogenerators. These are located in the 600 m (1,968 ft 6 in)-long machine hall, adjacent to the reactor building. The turbines, the venerable five-cylinder K-500-65/3000, are supplied by the Kharkiv turbine plant; the electrical generators are the TVV-500. The turbine and the generator rotors are mounted on the same shaft; the combined weight of the rotors is almost 200 t (220 short tons) and their nominal rotational speed is 3000 rpm. The turbogenerator is 39 m (127 ft 11 in) long and its total weight is 1,200 t (1,300 short tons). The coolant flow for each turbine is 82,880 t (91,360 short tons)/h. The generator produces 20 kV 50 Hz AC power. The generator's stator is cooled by water while its rotor is cooled by hydrogen. The hydrogen for the generators is manufactured on-site by electrolysis. The design and reliability of the turbines earned them the State Prize of Ukraine for 1979.
The Kharkiv turbine plant (now Turboatom) later developed a new version of the turbine, K-500-65/3000-2, in an attempt to reduce use of valuable metal. The Chernobyl plant was equipped with both types of turbines; Block 4 had the newer ones.
As an early Generation II reactor based on 1950s Soviet technology, the RBMK design was optimized for speed of production over redundancy. It was designed and constructed with several design characteristics that proved dangerously unstable when operated outside their design specifications. The decision to use a superheated, vacuum-isolated graphite core with natural uranium fuel allowed for massive power generation at only a quarter of the expense of heavy water reactors, which were more maintenance-intensive and required large volumes of expensive heavy water for startup. However, it also had unexpected negative consequences that would not reveal themselves fully until the 1986 Chernobyl disaster.
Light water (ordinary H2O) is both a neutron moderator and a neutron absorber. This means that not only can it slow down neutrons to velocities in equilibrium with surrounding molecules ("thermalize" them and turn them into low-energy neutrons, known as thermal neutrons, that are far more likely to interact with the uranium-235 nuclei than the fast neutrons produced by fission initially), but it also absorbs some of them.
In RBMKs, light water was used as a coolant; moderation was mainly carried out by graphite. As graphite already moderated neutrons, light water had a lesser effect in slowing them down, but could still absorb them. This means that the reactor's reactivity (adjustable by appropriate neutron-absorbing rods) had to account for the neutrons absorbed by light water.
In the case of evaporation of water to steam, the place occupied by water would be occupied by water vapor, which has a density vastly lower than that of liquid water (the exact number depends on pressure and temperature; at standard conditions, steam is about 1⁄1350 as dense as liquid water). Because of this lower density (of mass, and consequently of atom nuclei able to absorb neutrons), light water's neutron-absorption capability practically disappears when it boils. This allows more neutrons to fission more U-235 nuclei and thereby increase the reactor power, which leads to higher temperatures that boil even more water, creating a thermal feedback loop.
In RBMKs, generation of steam in the coolant water would then in practice create a void, a bubble that does not absorb neutrons; the reduction in moderation by light water is irrelevant, as graphite is still moderating the neutrons. However, the loss of absorption would dramatically alter the balance of neutron production, causing a runaway condition in which more and more neutrons are produced, and their density grows exponentially fast. Such a condition is called a positive void coefficient , and the RBMK has the highest positive void coefficient of any commercial reactor ever designed.
A high void coefficient does not necessarily make a reactor inherently unsafe, as some of the fission neutrons are emitted with a delay of seconds or even minutes (post-fission neutron emission from daughter nuclei), so steps can be taken to reduce the fission rate before it gets too high. However, it does make it considerably harder to control the reactor (especially at low power) and makes it imperative that the control systems are very reliable and the control room personnel (regardless of rank or position) are rigorously trained in the peculiarities and limits of the system. Neither of these requirements were in place at Chernobyl: since the reactor's actual design bore the approval stamp of the Kurchatov Institute and was considered a state secret, discussion of the reactor's flaws was forbidden, even among the actual personnel operating the plant. Some later RBMK designs did include control rods on electromagnetic grapples, thus controlling the reaction speed and, if necessary, stopping the reaction completely. The RBMK at Chernobyl, however, had manual control rods.
After the Chernobyl disaster, all RBMKs in operation underwent significant changes, lowering their void coefficients from +4.7 β to +0.7 β. This new number decreases the possibility of a low-coolant meltdown.
In his posthumously published memoirs, Valery Legasov, the First Deputy Director of the Kurchatov Institute of Atomic Energy, revealed that the Institute's scientists had long known that the RBMK had significant design flaws.Legasov's suicide in 1988, apparently a result of becoming bitterly disappointed with the failure of the authorities to confront the flaws, caused shockwaves throughout the Soviet nuclear industry and the problems with the RBMK design were rapidly accepted.
Following Legasov's death, all remaining RBMKs were retrofitted with a number of updates for safety. The largest of these updates fixed the RBMK control rod design. The control rods have 4.5 m (14 ft 9 in) graphite displacers, which prevent coolant water from entering the space vacated as the rods are withdrawn. In the original design, those displacers, being shorter than the height of the core, left 1.25 metres (4.1 ft) columns of water at the bottom (and 1.25 metres (4.1 ft) at the top) when the rods were fully extracted; during insertion, the graphite would first displace that lower water, locally increasing reactivity. Also, when the rods were in their uppermost position, the absorber ends were outside the core, requiring a relatively large displacement before achieving a significant reduction in reactivity.[ citation needed ] These design flaws were likely the final trigger of the first explosion of the Chernobyl accident, causing the lower part of the core to become supercritical when the operators tried to shut down the highly destabilized reactor by reinserting the rods.
The updates are:
In addition, RELAP5-3D models of RBMK-1500 reactors were developed for use in integrated thermal-hydraulics-neutronics calculations for the analysis of specific transients in which the neutronic response of the core is important.
From May 2012 to December 2013, Leningrad-1 was offline while repairs were made related to deformed graphite moderator blocks. The 18-month project included research and the development of maintenance machines and monitoring systems. Similar work will be applied to the remaining operational RBMKs.Graphite moderator blocks in the RBMK can be repaired and replaced in situ, unlike in the other current large graphite moderated reactor, the Advanced gas-cooled reactor.
Longitudinal cutting in some of the graphite columns during lifetime extension refurbishment work can return the graphite stack its initial design geometry.
A post-Soviet redesign of the RBMK is the MKER (Russian: МКЭР, Многопетлевой Канальный Энергетический Реактор [Mnogopetlevoy Kanalniy Energeticheskiy Reaktor] which means Multi-loop pressure tube power reactor), with improved safety and containment.The physical prototype of the MKER-1000 is the 5th unit of the Kursk Nuclear Power Plant. The construction of Kursk 5 was cancelled in 2012. A MKER-800, MKER-1000 and MKER-1500 are planned for the Leningrad nuclear power plant.
Of the 17 RBMKs built (one was still under construction at the Kursk Nuclear Power Plant), all three surviving reactors at the Chernobyl plant have now been closed (the fourth having been destroyed in the accident, and the second disabled after a hydrogen explosion in 1991). Chernobyl 5 and 6 were under construction at the time of the accident at Chernobyl, but further construction was stopped due to the high level of contamination at the site limiting its longer term future. Both reactors at Ignalina in Lithuania were also shut down.Russia is the only country to still operate reactors of this design: Leningrad (3 RBMK-1000), Smolensk (3 RBMK-1000) and Kursk (4 RBMK-1000).
|– Operational reactor (including reactors currently offline)||– Reactor decommissioned||– Reactor destroyed||– Abandoned or cancelled reactor|
|Chernobyl-1||RBMK-1000||shut down in 1996||740||800|
|Chernobyl-2||RBMK-1000||shut down in 1991||925||1,000|
|Chernobyl-3||RBMK-1000||shut down in 2000||925||1,000|
|Chernobyl-4||RBMK-1000||destroyed in the 1986 accident||925||1,000|
|Chernobyl-5||RBMK-1000||construction cancelled in 1988||950||1,000|
|Chernobyl-6||RBMK-1000||construction cancelled in 1988||950||1,000|
|Ignalina-1||RBMK-1500||shut down in 2004||1,185||1,300 [A]|
|Ignalina-2||RBMK-1500||shut down in 2009||1,185||1,300 [A]|
|Ignalina-3||RBMK-1500||construction cancelled in 1988||1,380||1,500|
|Ignalina-4||RBMK-1500||plan cancelled in 1988||1,380||1,500|
|Kostroma-1||RBMK-1500||construction cancelled in 1980s||1,380||1,500|
|Kostroma-2||RBMK-1500||construction cancelled in 1980s||1,380||1,500|
|Kursk-1||RBMK-1000||operational until 2022||925||1,000|
|Kursk-2||RBMK-1000||operational until 2024||925||1,000|
|Kursk-3||RBMK-1000||operational until 2029||925||1,000|
|Kursk-4||RBMK-1000||operational until 2030||925||1,000|
|Kursk-5||MKER-1000 [B]||construction cancelled in 2012||925||1,000|
|Kursk-6||RBMK-1000||construction cancelled in 1993||925||1,000|
|Leningrad-1||RBMK-1000||shut down in 2018||925||1,000|
|Leningrad-2||RBMK-1000||operational until 2021||925||1,000|
|Leningrad-3||RBMK-1000||operational until June 2025||925||1,000|
|Leningrad-4||RBMK-1000||operational until August 2026||925||1,000|
|Smolensk-1||RBMK-1000||operational until 2028||925||1,000|
|Smolensk-2||RBMK-1000||operational until 2030||925||1,000|
|Smolensk-3||RBMK-1000||operational until 2050||925||1,000|
|Smolensk-4||RBMK-1000||construction cancelled in 1993||925||1,000|
|A Built with 1,500 MWe gross electric power, the RBMK-1500 were de-rated to 1,360 MW after the Chernobyl disaster.|
|B Kursk-5 is the unfinished physical prototype for the MKER class of nuclear power plants, a once planned successor to the RBMK class of power plants. Kursk-5 features a MKER reactor core in a modified RBMK building. No MKER of any type has yet been completed.|
A graphite-moderated Magnox reactor similar to the RBMK design exists in North Korea at the Yongbyon Nuclear Scientific Research Center.
Pressurized water reactors (PWRs) constitute the large majority of the world's nuclear power plants and are one of three types of light-water reactor (LWR), the other types being boiling water reactors (BWRs) and supercritical water reactors (SCWRs). In a PWR, the primary coolant (water) is pumped under high pressure to the reactor core where it is heated by the energy released by the fission of atoms. The heated water then flows to a steam generator where it transfers its thermal energy to a secondary system where steam is generated and flows to turbines which, in turn, spin an electric generator. In contrast to a boiling water reactor, pressure in the primary coolant loop prevents the water from boiling within the reactor. All LWRs use ordinary water as both coolant and neutron moderator.
A boiling water reactor (BWR) is a type of light water nuclear reactor used for the generation of electrical power. It is the second most common type of electricity-generating nuclear reactor after the pressurized water reactor (PWR), which is also a type of light water nuclear reactor. The main difference between a BWR and PWR is that in a BWR, the reactor core heats water, which turns to steam and then drives a steam turbine. In a PWR, the reactor core heats water, which does not boil. This hot water then exchanges heat with a lower pressure water system, which turns to steam and drives the turbine. The BWR was developed by the Argonne National Laboratory and General Electric (GE) in the mid-1950s. The main present manufacturer is GE Hitachi Nuclear Energy, which specializes in the design and construction of this type of reactor.
An Advanced Gas-cooled Reactor (AGR) is a type of nuclear reactor designed and operated in the United Kingdom. These are the second generation of British gas-cooled reactors, using graphite as the neutron moderator and carbon dioxide as coolant. They have been the backbone of the UK's nuclear generation fleet since the 1980s.
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 coated ceramic layer 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.
A nuclear meltdown is a severe nuclear reactor accident that results in core damage from overheating. The term nuclear meltdown is not officially defined by the International Atomic Energy Agency or by the Nuclear Regulatory Commission. It has been defined to mean the accidental melting of the core of a nuclear reactor, however, and is in common usage a reference to the core's either complete or partial collapse.
The A2W reactor is a naval nuclear reactor used by the United States Navy to provide electricity generation and propulsion on warships. The A2W designation stands for:
A loss-of-coolant accident (LOCA) is a mode of failure for a nuclear reactor; if not managed effectively, the results of a LOCA could result in reactor core damage. Each nuclear plant's emergency core cooling system (ECCS) exists specifically to deal with a LOCA.
The light-water reactor (LWR) is a type of thermal-neutron reactor that uses normal water, as opposed to heavy water, as both its coolant and neutron moderator – furthermore a solid form of fissile elements is used as fuel. Thermal-neutron reactors are the most common type of nuclear reactor, and light-water reactors are the most common type of 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.
The Chernobyl disaster was a nuclear accident that occurred on 26 April 1986 at the No. 4 nuclear reactor in the Chernobyl Nuclear Power Plant, near the city of Pripyat in the north of the Ukrainian SSR. It is considered the worst nuclear disaster in history and is one of only two nuclear energy disasters rated at seven—the maximum severity—on the International Nuclear Event Scale, the other being the 2011 Fukushima Daiichi nuclear disaster in Japan.
The supercritical water reactor (SCWR) is a concept Generation IV reactor, mostly designed as light water reactor (LWR) that operates at supercritical pressure. The term critical in this context refers to the critical point of water, and must not be confused with the concept of criticality of the nuclear reactor.
A nuclear reactor core is the portion of a nuclear reactor containing the nuclear fuel components where the nuclear reactions take place and the heat is generated. Typically, the fuel will be low-enriched uranium contained in thousands of individual fuel pins. The core also contains structural components, the means to both moderate the neutrons and control the reaction, and the means to transfer the heat from the fuel to where it is required, outside the core.
Steam generators are heat exchangers used to convert water into steam from heat produced in a nuclear reactor core. They are used in pressurized water reactors (PWR) between the primary and secondary coolant loops.
The three primary objectives of nuclear reactor safety systems as defined by the U.S. Nuclear Regulatory Commission are to shut down the reactor, maintain it in a shutdown condition and prevent the release of radioactive material.
The MKER is a Russian third generation nuclear reactor design. It is a development of the RBMK nuclear power reactor.
A nuclear reactor coolant is a coolant in a nuclear reactor used to remove heat from the nuclear reactor core and transfer it to electrical generators and the environment. Frequently, a chain of two coolant loops are used because the primary coolant loop takes on short-term radioactivity from the reactor.
Boiling water reactor safety systems are nuclear safety systems constructed within boiling water reactors in order to prevent or mitigate environmental and health hazards in the event of accident or natural disaster.
The Integral Molten Salt Reactor (IMSR) is a design for a small modular reactor (SMR) that employs molten salt reactor technology 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, and 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.