RBMK - WikiHQ

The RBMK (, 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. It is somewhat like a boiling water reactor as water boils in the pressure tubes. It is one of two power reactor types to enter serial production in the Soviet Union during the 1970s, the other being the VVER reactor. The name refers to its design 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 with the aggregate power of operational units at almost 7 GW of installed capacity. 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 thirty years.

RBMK reactors may be classified as belonging to one of three distinct generations, according to when the particular reactor was built and brought online:

Generation 1 – during the early-to-mid 1970s, before OPB-82 General Safety Provisions were introduced in the Soviet Union.
Generation 2 – during the late 1970s and early 1980s, conforming to the OPB-82 standards issued in 1982.
Generation 3 – post Chernobyl accident in 1986, where Soviet safety standards were revised to OPB-88; only Smolensk-3 was built to these standards.

Beside RBMK, there are several other graphite-moderated reactors. A graphite-moderated Magnox reactor exists in North Korea at the Yongbyon Nuclear Scientific Research Center. While the gas cooled Magnox, AGR and pebble bed reactors (Such as the Dragon reactor at Winfrith) use graphite as moderators their use of gases (carbon dioxide for Magnox and AGR, while helium for Dragon) as heat transfer fluids causes them to have no void coefficient. 4 EGP-6 graphite water reactors which are a scaled down version of the RBMK were operating at the world's second northern most nuclear power plant i.e. the Bilibino Nuclear Power Plant. One EGP-6 was shut down permanently in 2020. The remaining three EGP-6 were shutdown permanently in December 2025.

Lifespan

Initially the service life of the RBMK reactor type was expected to be 30 years, but this may be extended to 45 years with mid-life refurbishments (such as fixing the issue of graphite stack deformation in the core), and eventually a 50-year lifetime was adopted for some units (Kursk 1-3 and 1-4, Leningrad 1-3 and 1-4, Smolensk 1-1, 1-2, 1-3). Efforts are underway to extend the license of all the units. In July 2024, Leningrad unit 3's license was extended from 2025 to 2030. In February 2026, the Russian nuclear regulator approved a five year life extension to Leningrad Unit 4 to operate until 2031. Today all the reactors are operated by Rosatom's subsidiary Rosenergoatom.

In 2026, it was reported that the RBMK units of the Kursk plant had undergone significant safety upgrades reducing risks by almost 100 times.

History

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. The first of these, Obninsk AM-1 ("Атом Мирный", Atom Mirny, Russian for "peaceful atom," analogous to the American 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.

The RBMK was a minimalist design that used regular (light) water for cooling and graphite for moderation, making it possible to use fuel with a lower enrichment (1.8% enriched uranium instead of considerably more expensive 4% enrichment). This allowed for an extraordinarily large and powerful reactor that could be built rapidly, largely out of parts fabricated on-site instead of by specialized factories. Because a containment building would have needed to be very large and expensive, doubling the cost of each unit, due to the large size of the RBMK, it was originally omitted from the design. It was argued by its designers that the RBMK's strategy of having each fuel assembly in its own channel with flowing cooling water was an acceptable alternative for containment.

The RBMK was mainly designed at the Kurchatov Institute of Atomic Energy and , headed by Anatoly Aleksandrov and Nikolai Dollezhal respectively, from 1964 to 1966. A top-secret invention patent for the RBMK design was filed by Aleksandrov, who personally took credit for the design of the reactor, with the Soviet patent office. The RBMK was proclaimed by some as the national reactor of the Soviet Union, probably due to nationalism because of its unique design, large size and power output. Meanwhile the VVER (an earlier Soviet reactor design) was called the "American reactor" due to the pressurized water (PWR) design shared with many Western reactors. The RBMK was favored over the VVER by the Soviet Union due to its ease of manufacture, due to a lack of a large and thick-walled reactor pressure vessel and relatively complex associated steam generators, and its large power output, which would allow the Soviet government to easily meet their central economic planning targets.

The RBMK-1000's design was finalized in 1968. At that time it was the world's largest nuclear reactor design, surpassing western designs and the VVER in power output and physical size, being 20 times larger by volume than contemporary western reactors. Similarly to CANDU reactors and the Indian IPHWR reactors it could be produced without the specialized industry required by the large and thick-walled reactor pressure vessels such as those used by VVER reactors, thus increasing the number of factories capable of manufacturing RBMK reactor components.

The initial 1000 MWe design of the RBMK-1000 left room for development into yet more powerful reactors. 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. For comparison, the EPR has a net electric nameplate capacity of 1600 MW (4500 MWthermal) and is among the most powerful reactor types ever built.

The flaws in the original RBMK design were recognized by others, including from within the Kurchatov Institute before the first units were built, but the orders for construction of the first RBMK units, which were at Leningrad, had already been issued in 1966 by the Soviet government by the time their concerns reached the Central Committee of the Communist Party of the Soviet Union and the Soviet Council of Ministers. This prompted a sudden overhaul of the RBMK. Plutonium production in an RBMK would have been achieved by operating the reactor under special thermal parameters, but this capability was abandoned early on. This was the design that was finalized in 1968. The redesign did not solve further flaws that were not discovered until years later.

No prototypes of the RBMK were built; it was put directly into mass production. Construction of the first RBMK, which was at Leningrad Nuclear Power Plant, began in 1970. Leningrad unit 1 opened in 1973.

At Leningrad it was discovered that the RBMK, due to its high positive void coefficient, became harder to control as the uranium fuel was consumed or burned up, becoming unpredictable by the time it was shut down after three years for maintenance. This made controlling the RBMK a very laborious, mentally and physically demanding task requiring the timely adjustment of dozens of parameters every minute, around the clock, constantly wearing out switches such as those used for the control rods and causing operators to sweat. The enrichment percentage was increased to 2.0%, up from 1.8% to alleviate these issues.

The RBMK was considered by some in the Soviet Union to be already obsolete shortly after the commissioning of Chernobyl unit 1 (completed in 1977). Aleksandrov and Dollezhal did not investigate further or even deeply understand the problems in the RBMK, and the void coefficient was not analyzed in the manuals for the reactor. Engineers at Chernobyl unit 1 had to create solutions to many of the RBMK's flaws such as a lack of protection against no feedwater supply. Leningrad and Chernobyl units 1 both had partial meltdowns that were — alongside other nuclear accidents at Soviet power plants — treated as state secrets, and so were unknown even to other workers at those same plants.

By 1980 NIKIET realized, after completing a confidential study, that accidents with the RBMK were likely even during normal operation, but no action was taken to correct the RBMK's flaws. Instead, manuals were revised, which was believed to be enough to ensure safe operation as long as they were followed closely. However, the manuals were vague and Soviet power plant staff already had a habit of bending the rules to meet economic targets in the face of inadequate or malfunctioning equipment. Crucially, it was not made clear that a minimum number of control rods had to stay in the reactor at all times in order to protect against an accident, as loosely articulated by the Operational Reactivity Margin (ORM) parameter. An ORM chart recorder and display were added to RBMK control rooms after the Chernobyl disaster.

Reactor design and performance

Reactor vessel, moderator and shielding

thumb|upright=1.5|Schematic diagram of an RBMK

thumb|upright=1.5|Schematic side view of the layout of an RBMK reactor core

thumb|upright=1.5| The reactor hall and piping systems of the RBMK reactor.

The reactor pit or vault is made of reinforced concrete and has dimensions 21.6 m × 21.6 m × 25.5 m. It houses the vessel of the reactor, which is annular, made of an inner and outer cylindrical wall and top and bottom metal plates that cover the space between the inner and outer walls, without covering the space surrounded by the vessel. The reactor vessel is an annular steel cylinder with hollow walls and pressurized with nitrogen gas, with an inner diameter and height of 14.52 m × 9.7 m, and a wall thickness of 16 mm.

In order to absorb axial thermal expansion loads, it is equipped with two annular bellows compensators, one on the top and another on the bottom, in the spaces between the inner and outer walls. The vessel surrounds the graphite core block stack, which serves as moderator. The graphite stack is kept in a helium-nitrogen mixture, providing an inert atmosphere for the graphite, protecting it from potential fires, and facilitating transfer of excess heat from the graphite to the coolant channels.

The moderator blocks are made of nuclear graphite, the dimensions of which are 25 cm × 25 cm on the plane perpendicular to the channels, and with several longitudinal dimensions of between 20 cm and 60 cm depending on the location in the stack. There are holes of 11.4 cm diameter through the longitudinal axis of the blocks for the fuel and control channels. The blocks are stacked, surrounded by the reactor vessel into a cylindrical core with a diameter and height of 14 m × 8 m. a welded structure with 3 cm thick walls, an inner diameter of 16.6 m and an outer diameter of 19 m, 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.

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 through the gas circuit.

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 online refueling machine.

The fuel channels may, instead of fuel, contain fixed neutron absorbers, or be filled completely with cooling water. They may also contain silicon-filled tubes in place of a fuel assembly, for the purpose of doping for semiconductors. These channels could be identified by their corresponding servo readers, which would be blocked and replaced with the atomic symbol for silicon.

The small clearance between the pressure channel and the graphite block makes the graphite core susceptible to damage. If a pressure channel deforms, for example, by too high an internal pressure, the deformation can cause significant pressure loads on the graphite blocks and lead to damage.

Fuel

thumb|RBMK reactor fuel rod holder 1 – distancing armature; 2 – fuel rods shell; 3 – fuel tablets.

thumb|RBMK reactor fuel rod holder Uranium fuel pellets, fuel tubes, distancing armature, graphite bricks.

The fuel pellets are made of uranium dioxide powder, sintered with a suitable binder into pellets 11.5 mm in diameter and 15 mm 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 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 in 1980 was 2% (0.4% for the end pellets of the assemblies). Maximum allowable temperature of the fuel pellet is 2100 °C.

The fuel rods are zircaloy (1% niobium) tubes 13.6 mm in outer diameter, 0.825 mm 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 of fuel pellets. The fuel rods are 3.64 m long, with 3.4 m of that being the active length. The maximum allowed temperature of a fuel rod is 600 °C.

Unlike the rectangular PWR/BWR fuel assemblies or hexagonal VVER 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. 78 seconds before the reactor exploded.]]

The majority 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 long graphite section at the end, separated by a 1.25 m 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 neutron-absorbing water as the rod moves down could cause 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.

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 reactor control and protection system 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 channels by a thin layer of water between an inner jacket and the Zircaloy tube of the channel 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.

Gas circuit

The reactor operates in a helium–nitrogen atmosphere (70–90% He, 10–30% N2 by volume). 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 space between reactors in second generation RBMKs such as Kursk and Chernobyl 3/4 or some distance away from the reactors in first generation RBMKs such as Kursk and Chernobyl 1/2.

The gas is injected to the core 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. There is an ion exchange system included in the loop to remove impurities from the feedwater.

The turbine consists of one high-pressure rotor (cylinder) 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 (ion-exchange) 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 6,000–7,000 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 /h.

The nominal temperature of the coolant at the inlet of the reactor is about and the outlet temperature , at pressure in the drum separator and reactor of .

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 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.

In RBMK control rooms there are two large panels or mimic displays representing a top view of the reactor. One display is made up mostly or completely (in first generation RBMKs) of colored dials or rod position indicators: these dials represent the position of the control rods inside the reactor and the color of the housing of the dials matches that of the control rods, whose colors correspond to their function, for example, red for automatic control rods. The other display is a core map or core channel cartogram and is circular, is made of tiles, and represents every channel on the reactor. Each tile is made of a single light cover with a channel number and an incandescent light bulb, and each light bulb illuminates to represent out-of-spec (higher or lower than normal) channel parameters.

Operators have to type in the number of the affected channel(s) and then view the instruments to find exactly what parameters are out of spec. The core map represented information from the SKALA computer. Each unit had its own computer housed in a separate room. The control room also has chart or trend recorders. Some RBMK control rooms have been upgraded with video walls that replace the mimic displays and most chart recorders and eliminate the need to type in channel numbers and instead operators lay a cursor over a (now representative) tile to reveal its parameters that are shown on the lower side of the video wall. The control room is located below the floor of the deaerator room. Both rooms are in the space between the reactor and turbine buildings.

According to Unit 3 of Smolensk NPP, the reactor shutdown consists of (1) system of fast-acting emergency protection (BAZ) and (2) system of emergency protection (AZ-1). The BAZ has normally-withdrawn 24 fast acting scram rod. Meanwhile, AZ-1 has 32 bottom rods (bottom-inserted rods for axial power shaping), 9 local power regulation rods (automatically flux monitor-controlled for 9 local control zones), 18 local protection rods (2 rods for each local control zones, inserted if local power go beyond 10% preset or go beyond 2% preset accompanied by AZ-3 or AZ-4 modes), and 128 manual control rods.

Containment

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. 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, and the Indian IPHWR reactor, both for refueling and for plutonium production for nuclear weapons. This required large cranes above the core.

As the RBMK reactor core is very tall (about ), 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 core. 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 of the older RBMK reactors were retrofitted with an accident containment system, akin to that boasted by Chernobyl Unit 4.

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", not to be confused with the Russian ELENA reactor), sever the rest of the fuel channels, destroy the control rod insertion system, and potentially also withdraw control rods from the core.

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 . The distribution headers and inlets enclosures can handle and are vented via check valves to the leaktight compartment. The reactor cavity can handle overpressure of 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.

Turbogenerators

The electrical energy is generated by a pair of 500 MW hydrogen-cooled turbogenerators. These are located in the -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 and their nominal rotational speed is 3000 rpm. RBMK-1500 has increased coolant flow of the core. Because of the RBMK's positive void coefficient, the reduced cooling water volume causes a higher power output. As the name suggests, it was designed for an electrical power output of 1500 MW. The only reactors of this type and power output are the ones at Ignalina Nuclear Power Plant. and RBMK-3600 were designed to produce 2000 and 3600 MW of electrical power respectively. The RBMK-2000 would have had an increased channel diameter and number of fuel rods per fuel assembly while maintaining the same dimensions of the reactor core as the RBMK-1000 and RBMK-1500. The RBMK-3600 presumably similarly to the RBMK-1500 would have added turbulators to the RBMK-2000 design to increase heat removal.

RBMKP-2400

thumb|RBMKP-2400 reactor channel

The () is rectangular instead of cylindrical, and it was a modular, theoretically infinitely longitudinally expandable design with vertical steam separators, intended to be made in sections at a factory for assembly in situ. It was designed to have a power output of 2400 MWe, and a higher thermal efficiency due to steam superheating directly in the reactor core in special fuel channels with fuel rods with stainless steel cladding instead of the more common Zircaloy cladding, for a steam outlet temperature of 450 °C. 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 development of this design was cancelled in the aftermath of the Chernobyl disaster. An RBMKP-4800 would have had an increased number of evaporating and superheating channels thus increasing power output. Two RBMKP-2400s were planned for the Kostroma Nuclear Power Plant.

Technical specifications

{| class="wikitable"

|+Technical specifications of RBMK

!Characteristics

!RBMK-1000

!RBMK-1500

!RBMK-2000

!RBMK-3600

!RBMKP-2400

|References

|Thermal power of the reactor, MW

|3200

|4800

|5400

|Electrical capacity of the unit, MW

|1000

|1500

|2000

|3600

|2400

|Unit efficiency (gross), %

|31.25

|37.04

|Steam pressure in front of the turbine, atm

|65

|Steam temperature in front of the turbine, °C

|280

|450

|Dimensions active zone, m:

|— height

|7.05

|— diameter (width×length)

|11.8

|7.05×25.38

|Uranium loading, t

|192

|189

|220

|Enrichment, % 235U

|— evaporation channel

|2.6-3.0

|2.6-2.8

|1.8

|— superheating channel

|2.2

|Number of channels:

|— evaporative

|22.5

|25.4

|20.2

|— superheating

|18.9

|Average burnup, MWd/kg

|— in the evaporation channel

|22.5

|25.4

|20.2

|— in the superheating duct

|18.9

|Fuel shell dimensions (diameter×thickness), mm:

| — evaporation channel

|13.5×0.9

| — superheating channel

|10×0.3

|Fuel element shell material:

| — evaporation channel

|Zr + 2.5% Nb

| — superheating channel

|Stainless steel

|Number of fuel rods in the cassette

|18

|Number of cassettes

|1693

|1661

Terminology

{| class="wikitable sortable"

|+Terminology in RBMK

! colspan="3" |Russian

! rowspan="2" |English

! rowspan="2" |Note

!Initialism

!Full term

!Latin transcription

|АВР

|автоматический ввод резерва

|avtomaticheskiy vvod rezerva

|automatic transfer switch

|АЗ

|аварийная защита

|avariynaya zashchita

|emergency protection

|АЗ

|активная зона

|aktivnaya zona

|nuclear reactor core

|АЗ-1

|аварийная защита 1

|avariynaya zashchita 1

|emergency protection 1

|reducing the reactor power to 60% of the nominal power

|АЗ-2

|аварийная защита 2

|avariynaya zashchita 2

|emergency protection 2

|reducing the reactor power to 50% of the nominal power

|АЗ-5

|аварийная защита 5

|avariynaya zashchita 5

|emergency protection 5 (scram)

|emergency protection system installed in power units with RBMK reactors

|АЗМ

|аварийная защита (сигнал) по превышению мощности

|avariynaya zashchita (signal) po prevysheniyu moshchnosti

|emergency protection (signal) for excess power

|АЗММ

|аварийная защита (сигнал) по диапазону малой мощности

|avariynaya zashchita (signal) po diapazonu maloy moshchnosti

|emergency protection (signal) for low power range

|АЗРТ

|аварийная защита реакторной установки по технологическим параметрам (система)

|avariynaya zashchita reaktornoy ustanovki po tekhnologicheskim parametram (sistema)

|emergency protection of a reactor plant based on technological parameters (system)

|АЗС

|аварийная защита (сигнал) по высокой температуре

|avariynaya zashchita (signal) po vysokoy temperature

|high temperature alarm

|АЗС-П

|аварийная защита по аварийному увеличению скорости нарастания мощности в пусковом диапазоне

|avariynaya zashchita po avariynomu uvelicheniyu skorosti narastaniya moshchnosti v puskovom diapazone

|emergency protection for emergency increase in the rate of power increase in the starting range

|АЗС-Р

|аварийная защита по скорости в рабочем диапазоне мощности реактора

|avariynaya zashchita po skorosti v rabochem diapazone moshchnosti reaktora

|emergency protection for speed in the operating range of reactor power

|АИС

|автоматизированная измерительная система

|avtomatizirovannaya izmeritel'naya sistema

|automated measuring system

|АПН

|аварийный питательный насос

|avariynyy pitatel'nyy nasos

|emergency feed pump

|АР

|автоматический регулятор

|avtomaticheskiy regulyator

|automatic regulator

|АСКРО

|автоматизированная система контроля радиационной обстановки

|avtomatizirovannaya sistema kontrolya radiatsionnoy obstanovki

|automated radiation monitoring system

|АСУТП

|автоматизированная система управления технологическими процессами

|avtomatizirovannaya sistema upravleniya tekhnologicheskimi protsessami

|automated process control system

|БАЗ

|быстродействующая аварийная защита

|bystrodeystvuyushchaya avariynaya zashchita

|fast-acting emergency protection

|BAZ

|ББ

|бассейн-барботер

|basseyn-barboter

|bubbler pool

|БИК

|боковая ионизационная камера

|bokovaya ionizatsionnaya kamera

|side ionization chamber

|БОУ

|блочная обессоливающая установка

|blochnaya obessolivayushchaya ustanovka

|block desalination plant

|БПВ

|бак питательной воды

|bak pitatel'noy vody

|feedwater tank

|БПУ

|блочная панель управления

|blochnaya panel' upravleniya

|block control panel

|БРУ-Б

|быстродействующее редукционное устройство со сбросом в барботер

|bystrodeystvuyushcheye reduktsionnoye ustroystvo so sbrosom v barboter

|steam relief valve for steam discharge to accident localization system tower (bubbler)

|БРУ-Д

|быстродействующее редукционное устройство со сбросом в деаэратор

|bystrodeystvuyushcheye reduktsionnoye ustroystvo so sbrosom v deaerator

|steam relief valve for steam discharge to deaerator

|БРУ-К

|быстродействующее редукционное устройство со сбросом в конденсатор турбины

|bystrodeystvuyushcheye reduktsionnoye ustroystvo so sbrosom v kondensator turbiny

|steam relief valve for steam discharge to turbine condenser

|БРУ-ТК

|быстродействующее редукционное устройство со сбросом в технологический конденсатор

|bystrodeystvuyushcheye reduktsionnoye ustroystvo so sbrosom v tekhnologicheskiy kondensator

|high-speed pressure reducing device with discharge into a process condenser

|БС

|барабан-сепаратор

|baraban-separator

|drum separator

|БСМ

|быстрое снижение мощности

|bystroye snizheniye moshchnosti

|rapid decrease in power

|БЩУ

|блочный щит управления

|blochnyy shchit upravleniya

|block control panel

|БЩУ-Н

|блочный щит управления неоперативный

|blochnyy shchit upravleniya neoperativnyy

|non-operational block control panel

|БЩУ-О

|блочный щит управления оперативный

|blochnyy shchit upravleniya operativnyy

|operational block control panel

|ВЗД

|внутризонный датчик

|vnutrizonnyy datchik

|intra-zone sensor

|ВИК

|высотная ионизационная камера

|vysotnaya ionizatsionnaya kamera

|high-altitude ionization chamber

|ВК

|верхний концевой выключатель

|verkhniy kontsevoy vyklyuchatel'

|upper limit switch

|ВРД-В

|внутриреакторный датчик (контроля энерговыделения) высотный

|vnutrireaktornyy datchik (kontrolya energovydeleniya) vysotnyy

|in-reactor sensor (energy release monitoring) altitude

|ВРД-Р

|внутриреакторный датчик (контроля энерговыделения) радиальный

|vnutrireaktornyy datchik (kontrolya energovydeleniya) radial'nyy

|in-reactor sensor (power release monitoring) radial

|ВСРО

|вспомогательные системы реакторного отделения

|vspomogatel'nyye sistemy reaktornogo otdeleniya

|auxiliary systems of the reactor compartment

|ГПК

|главный предохранительный клапан

|glavnyy predokhranitel'nyy klapan

|main relief valve

|ГЦК

|главный циркуляционный контур

|glavnyy tsirkulyatsionnyy kontur

|main circulation circuit

|ГЦН

|главный циркуляционный насос

|glavnyy tsirkulyatsionnyy nasos

|main circulation pump

|MCP

|ДКЭ

|датчик контроля энерговыделения

|datchik kontrolya energovydeleniya

|energy release monitoring sensor

|ДП

|дополнительный поглотитель

|dopolnitel'nyy poglotitel'

|additional absorber

|ДРЕГ

|диагностическая регистрация параметров

|diagnosticheskaya registratsiya parametrov

|diagnostic recording of parameters

|DREG

|ДРК

|дроссельно-регулирующий клапан

|drossel'no-reguliruyushchiy klapan

|throttle control valve

|ДЭ

|деаэраторная этажерка

|deaeratornaya etazherka

|deaerator rack

|ЖРО

|жидкие радиоактивные отходы

|zhidkiye radioaktivnyye otkhody

|liquid radioactive waste

|ЗРК

|запорно-регулирующий клапан

|zaporno-reguliruyushchiy klapan

|shut-off and control valve

|ИПУ

|импульсное предохранительное устройство

|impul'snoye predokhranitel'noye ustroystvo

|impulse safety device

|ИСС

|информационно-измерительная система

|informatsionno-izmeritel'naya sistema

|information and measuring system

|КГО

|контроль герметичности оболочки

|kontrol' germetichnosti obolochki

|shell tightness control

|КД

|камера деления

|kamera deleniya

|fission chamber

|КИУМ

|коэффициент использования установленной мощности

|koeffitsiyent ispol'zovaniya ustanovlennoy moshchnosti

|installed capacity utilization factor

|КМПЦ

|контур многократной принудительной циркуляции

|kontur mnogokratnoy prinuditel'noy tsirkulyatsii

|multiple forced circulation circuit

|КОО

|канал охлаждения отражателя

|kanal okhlazhdeniya otrazhatelya

|reflector cooling channel

|КПР

|капитально-плановый ремонт

|kapital'no-planovyy remont

|major scheduled repairs

|КРО

|кластерный регулирующий клапан

|klasternyy reguliruyushchiy klapan

|cluster control valve

|КУС

|ключ управления стержнями

|klyuch upravleniya sterzhnyami

|rod control key

|КЦТК

|контроль целостности технологических каналов

|kontrol' tselostnosti tekhnologicheskikh kanalov

|control of the integrity of process channels

|ЛАЗ

|локальная аварийная защита

|lokal'naya avariynaya zashchita

|local emergency protection

|ЛАР

|локальный автоматический регулятор

|lokal'nyy avtomaticheskiy regulyator

|local automatic regulator

|МЗР

|максимальный запас реактивности

|maksimal'nyy zapas reaktivnosti

|maximum reactivity margin

|МПА

|максимальная проектная авария

|maksimal'naya proyektnaya avariya

|maximum design basis accident

|МТК

|мнемотабло технологических каналов

|mnemotablo tekhnologicheskikh kanalov

|mnemonic display of technological channels

|МФК

|минимальный физический уровень мощности

|minimal'nyy fizicheskiy uroven' moshchnosti

|minimum physical power level

|НВК

|нижние водяные коммуникации

|nizhniye vodyanyye kommunikatsii

|lower water communications

|НК

|напорный коллектор

|napornyy kollektor

|pressure manifold

|НСБ

|начальник смены блока

|nachal'nik smeny bloka

|shift supervisor

|НСС

|начальник смены станции

|nachal'nik smeny stantsii

|station shift supervisor

|НФХ

|нейтронно-физические характеристики

|neytronno-fizicheskiye kharakteristiki

|neutron-physical characteristics

|ОЗР

|оперативный запас реактивности

|operativnyy zapas reaktivnosti

|operating reactivity margin

|ORM

|ОК

|обратный клапан

|obratnyy klapan

|check valve

|ОПБ

|Общие положения безопасности

|obshchiye polozheniya bezopasnosti

|general safety provisions

|ПВД

|подогреватель высокого давления

|podogrevatel' vysokogo davleniya

|high pressure heater

|ПВК

|пароводяные коммуникации

|parovodyanyye kommunikatsii

|steam and water communications

|ПК-АЗ

|режим действия группы стержней перекомпенсации

|rezhim deystviya gruppy sterzhney perekompensatsii

|operating mode of the overcompensation rod group

|ПКД

|паровой компенсатор давления

|parovoy kompensator davleniya

|steam pressure compensator

|ПН

|питательный насос

|pitatel'nyy nasos

|feedwater pump

|ППБ

|плотно-прочный бокс

|plotno-prochnyy boks

|tightly-durable box

|ППР

|планово-предупредительный ремонт

|planovo-predupreditel'nyy remont

|scheduled preventive maintenance

|ПРИЗМА

|программа измерения мощности аппарата

|programma izmereniya moshchnosti apparata

|program for measuring the device's power

|PRIZMA, PRISMA

|ПСУ

|пассивное спринклерное устройство

|passivnoye sprinklernoye ustroystvo

|passive sprinkler device

|ПЭН

|питательный электронасос

|pitatel'nyy elektronasos

|feed electric pump

|ПЯБ

|Правила ядерной безопасности

|pravila yadernoy bezopasnosti

|Nuclear safety regulations

|РВ

|резервное возбуждение турбины

|rezervnoye vozbuzhdeniye turbiny

|turbine backup excitation

|РГК

|раздаточно-групповой коллектор

|razdatochno-gruppovoy kollektor

|distribution and group collector

|РЗМ

|разгрузочно-загрузочная машина

|razgruzochno-zagruzochnaya mashina

|loading and unloading machine

|РЗК

|разгрузочно-загрузочный комплекс

|razgruzochno-zagruzochnyy kompleks

|loading and unloading complex

|РК СУЗ

|рабочий канал системы управления и защиты

|rabochiy kanal sistemy upravleniya i zashchity

|working channel of the control and protection system

|РП

|реакторное пространство

|reaktornoye prostranstvo

|reactor space

|РР

|ручное регулирование

|ruchnoye regulirovaniye

|manual control

|РУ

|реакторная установка

|reaktornaya ustanovka

|reactor installation

|САОР

|система аварийного охлаждения реактора

|sistema avariynogo okhlazhdeniya reaktora

|emergency reactor cooling system

|СБ

|системы безопасности

|sistemy bezopasnosti

|security systems

|СВП

|стержень выгорающего поглотителя

|sterzhen' vygorayushchego poglotitelya

|burnable absorber rod

|СГО

|система герметичного ограждения

|sistema germetichnogo ograzhdeniya

|hermetic enclosure system

|СЛА

|система локализации аварий

|sistema lokalizatsii avariy

|accident localization system

|ALS

|Скала

|система контроля аппарата Ленинградской атомная электростанция

|sistema kontrolya apparata Leningradskoy atomnaya elektrostantsiya

|Leningrad NPP apparatus control system

|SKALA, SCALA

|СП

|стержень-поглотитель

|sterzhen'-poglotitel'

|absorber rod

|СПИР

|система пассивного отвода тепла

|sistema passivnogo otvoda tepla

|passive heat dissipation system

|СРК

|стопорно-регулирующий клапан

|stoporno-reguliruyushchiy klapan

|shut-off and control valve

|СТК

|система технологического контроля

|sistema tekhnologicheskogo kontrolya

|process control system

|СУЗ

|система управления и защиты

|sistema upravleniya i zashchity

|control and protection system

|SUZ

|СФКРЭ

|система физического контроля распределения энерговыделения

|sistema fizicheskogo kontrolya raspredeleniya energovydeleniya

|physical control system for energy distribution

|СЦК

|система централизованного контроля

|sistema tsentralizovannogo kontrolya

|centralized control system

|refers to SKALA

|ТВС

|тепловыделяющая сборка

|teplovydelyayushchaya sborka

|fuel assembly

|ТВЭЛ

|тепловыделяющий элемент

|teplovydelyayushchiy element

|fuel element, nuclear fuel

|TVEL

|ТГ

|турбогенератор

|turbogenerator

|turbo generator

|ТК

|технологический канал

|tekhnologicheskiy kanal

|technological channel

|ТН

|теплоноситель

|teplonositel'

|coolant

|УЗСП

|усилитель защиты по скорости пускового диапазона

|usilitel' zashchity po skorosti puskovogo diapazona

|starting range speed protection amplifier

|УСП

|укороченный стержень-поглотитель (ручной)

|ukorochennyy sterzhen'-poglotitel' (ruchnoy)

|shortened absorber rod (manual)

|USP

|УТЦ

|учебно-тренировочный центр

|uchebno-trenirovochnyy tsentr

|training center

|ЯТ

|ядерное топливо

|yadernoye toplivo

|nuclear fuel

|ЯТЦ

|ядерный топливный цикл

|yadernyy toplivnyy tsikl

|nuclear fuel cycle

|ЯЭУ

|ядерная энергетическая установка

|yadernaya energeticheskaya ustanovka

|nuclear power plant

Issues of RBMK

thumb|The 4th RBMK reactor of the [[Chernobyl Nuclear Power Plant, destroyed in the 1986 Chernobyl disaster.]]

Design flaws and safety issues

As an early Generation II reactor based on 1950s Soviet technology, the RBMK design was optimized for speed of production but sacrificed redundancy. Several of its design characteristics would prove to be dangerously unstable when operated outside their design specifications. The decision to use a 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, its unintended consequences would not reveal themselves fully until the Chernobyl disaster in 1986.

High positive void coefficient

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 the RBMK series of reactors, light water functions as a coolant, while moderation is mainly carried out by graphite. As graphite already moderates neutrons, light water has 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) must take into account the neutrons absorbed by light water.

In the case of vaporisation 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 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 RBMK reactors, 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 still moderates the neutrons. However, the loss of absorption dramatically alters the balance of neutron production, causing a runaway condition in which more and more neutrons are produced, and their density grows exponentially. Such a condition is called a "positive void coefficient", and the RBMK reactor series 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), and therefore steps can be taken to reduce the fission rate before it becomes too high. This situation, however, does make it considerably harder to control the reactor, especially at low power. Thus, control systems must be very reliable and control-room personnel must be rigorously trained in the peculiarities and limits of the system. Neither of these requirements was 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 reactor at Chernobyl, however, had manual clutch control rods.

All RBMK reactors underwent significant changes following the Chernobyl disaster. The positive void coefficient was reduced from +4.5 β to +0.7 β, decreasing the likelihood of further reactivity accidents, at the cost of higher enrichment requirements of the uranium fuel.

Improvements since the Chernobyl accident

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, following frustrated attempts to promote nuclear and industrial safety reform, caused shockwaves throughout the scientific community. The RBMK's design problems were discussed increasingly openly.

Following the accident at Chernobyl, all remaining RBMK reactors were retrofitted with a number of updates for safety. The largest of these updates fixed the RBMK control rod design. The control rods have 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 columns of water at the bottom (and at the top) when the rods were fully extracted. 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 prompt critical when the operators tried to shut down the highly destabilized reactor by reinserting the rods. The updates are:

An increase in fuel enrichment from 2% to 2.4% to compensate for control rod modifications and the introduction of additional absorbers.
Manual control rod count increased from 30 to 45.
80 additional absorbers inhibit operation at low power, where the RBMK design is most dangerous.
AZ-5 (emergency reactor shutdown or Scram) sequence reduced from 18 to 12 seconds.
Addition of the БАЗ or BAZ system, (rapid reactor emergency protection) which would insert 24 uniformly distributed rods into the reactor core via a modified drive mechanism within 1.8 to 2.5 seconds.
Precautions against unauthorized access to emergency safety systems.

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.

BAZ button is intended as a preemptive measure to bring down reactivity before AZ-5 is activated, to enable the safe and stable emergency shutdown of a RBMK.

Deformed graphite moderator blocks

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 , 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 to its initial design geometry.

Modernization and decommissioning RBMK

Due to growing demand of energy in Russia, RBMK reactors has been undergoing service life extension by modernisation. Originally, RBMK designed for 25 years of service life. The modernization includes rectification of problem related to the Chernobyl accident, ratification of current safety standards, and ultimately extension of service life. Because of said modernization, RBMKs achieve improvement on safety, reliability and economic efficiency.

Despite such feasibility for extensive modifications and upgrade, Soviet-designed reactors were closed.

Further development

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 a containment building. A MKER-800, MKER-1000 and MKER-1500 were planned for the Leningrad nuclear power plant.

List of RBMK reactors

Color key:

:{|

| – Operational reactor (including reactors currently offline)

| – Reactor decommissioned

| – Reactor destroyed in accident

| – Abandoned or cancelled reactor construction

{| class="wikitable sortable"

!style="width=22%; background-color:#CFCFCF;" |Location

!style="width=22%; background-color:#CFCFCF;" |Current Country

!style="width=88%; background-color:#CFCFCF;" |Reactor type

!style="width=22%; background-color:#CFCFCF;" |Online

!style="width=08%; background-color:#CFCFCF;" |Status

!style="width=08%; background-color:#CFCFCF;" |Net Capacity (MWe)

!style="width=08%; background-color:#CFCFCF;" |Gross Capacity (MWe)

|- style="background-color:#ffcbcb"

|Chernobyl-1

| Ukraine

|RBMK-1000

|1977

|shut down in 1996

|align="right" | 740

|align="right" | 800

|- style="background-color: #ffcbcb"

|Chernobyl-2

| Ukraine

|RBMK-1000

|1978

|shut down in 1991 due to turbine fire

|align="right" | 925

|align="right" | 1,000

|- style="background-color: #ffcbcb"

|Chernobyl-3

| Ukraine

|RBMK-1000

|1981

|shut down in 2000

|align="right" | 925

|align="right" | 1,000

|- style="background-color: #ffff03"

|Chernobyl-4

| Ukraine

|RBMK-1000

|1983

|destroyed in 1986

|align="right" | 925

|align="right" | 1,000

|- style="background-color: #99ccff"

|Chernobyl-5

| Ukraine

|RBMK-1000

|N/A

|construction cancelled in 1988

|align="right" | 925

|align="right" | 1,000

|- style="background-color: #99ccff"

|Chernobyl-6

| Ukraine

|RBMK-1000

|N/A

|construction cancelled in 1988

|align="right" | 925

|align="right" | 1,000

|- style="background-color: #ffcbcb"

|Ignalina-1

| Lithuania

|RBMK-1500

|1983

|shut down in 2004

|align="right" | 1,185

|align="right" | 1,300

|- style="background-color: #ffcbcb"

|Ignalina-2

| Lithuania

|RBMK-1500

|1987

|shut down in 2009

|align="right" | 1,185

|align="right" | 1,300

|- style="background-color: #99ccff"

|Ignalina-3

| Lithuania

|RBMK-1500

|N/A

|construction cancelled in 1988

|align="right" | 1,380

|align="right" | 1,500

|- style="background-color: #99ccff"

|Ignalina-4

| Lithuania

|RBMK-1500

|N/A

|plan cancelled in 1988

|align="right" | 1,380

|align="right" | 1,500

|- style="background-color: #99ccff"

|Kostroma-1

| Russia

|RBMKP-2400

|N/A

|construction cancelled in 1980s

|align="right" | 2,260

|align="right" | 2,400

|- style="background-color: #99ccff"

|Kostroma-2

| Russia

|RBMKP-2400

|N/A

|construction cancelled in 1980s

|align="right" | 2,260

|align="right" | 2,400

|- style="background-color: #ffcbcb"

|Kursk-1

| Russia

|RBMK-1000

|1977

|shut down in 2021

|align="right" | 925

|align="right" | 1,000

|- style="background-color: #ffcbcb

|Kursk-2

| Russia

|RBMK-1000

|1979

|shut down in 2024

|align="right" | 925

|align="right" | 1,000

|- style="background-color: #b9ffc5"

|Kursk-3

| Russia

|RBMK-1000

|1984

|operational until 2033

|align="right" | 925

|align="right" | 1,000

|- style="background-color: #ffcbcb"

|Leningrad-2

| Russia

|RBMK-1000

|1976

|shut down in 2020

|align="right" | 925

|align="right" | 1,000

|- style="background-color: #b9ffc5"

|Leningrad-3

| Russia

|RBMK-1000

|1979

|operational until 2030 (extended by 5 years in 2025)

|align="right" | 925

|align="right" | 1,000

|- style="background-color: #b9ffc5"

|Leningrad-4

| Russia

|RBMK-1000

|1981

|operational until 2031 (extended by 5 years in 2026)

|align="right" | 925

|align="right" | 1,000

|- style="background-color: #b9ffc5"

|Smolensk-1

| Russia

|RBMK-1000

|1983

|operational until 2028 According to the automated radiation control system, the radiation situation at the plant and in its monitoring zone was normal.

References

Sources and external links

Technical data on RBMK-1500 reactor at Ignalina nuclear power plant – a decommissioned RBMK reactor.
Chernobyl – A Canadian Perspective – A brochure describing nuclear reactors in general and the RBMK design in particular, focusing on the safety differences between them and CANDU reactors. Published by Atomic Energy of Canada Limited.