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RBMK Reactors – Appendix to Nuclear Power Reactors

(Updated February 2022)

• The RBMK is an unusual reactor pattern, 1 of two to emerge in the Soviet Union in the 1970s.
• The blueprint had several shortcomings, and was the design involved in the 1986 Chernobyl disaster.
• Major modifications have been made to the RBMK reactors still operating.

The Soviet-designed RBMK (reaktor bolshoy moshchnosty kanalny, loftier-power aqueduct reactor) is a water-cooled reactor with individual fuel channels and using graphite as its moderator. It is also known as the low-cal water graphite reactor (LWGR). As with a humid h2o reactor (BWR), h2o boils in the fuel channels (at about vi.ix MPa) and steam is separated above them in a unmarried circuit. It was designed over 1964-66 and is very different from most other power reactors. Its precursors were an experimental xxx MWt (v MWe) LWGR at Obninsk which started up in 1954, and 2 small epitome LWGR (AMB-100 & 200) units – Beloyarsk 1&2, which ran from 1964 and 1968 respectively. (The ADE reactors at Zheleznogorsk and Seversk used for plutonium production are similar to the RBMK simply with much lower power density and smaller fuel elements.)

The combination of graphite moderator and water coolant is found in no other power reactors in the earth. As the Chernobyl accident showed, several of the RBMK'due south pattern characteristics – in particular, the control rod design and a positive void coefficient – were unsafe. A number of significant design changes were made later on the Chernobyl accident to accost these problems.

Features of the RBMK

Fuel

Pellets of slightly-enriched uranium oxide are enclosed in a zircaloy tube three.65m long, forming a fuel rod. A set up of eighteen fuel rods is bundled cylindrically in a wagon to grade a fuel assembly. Two of these end on end occupy each pressure level tube.

Pressure tubes

Within the reactor each fuel associates is positioned in its own vertical force per unit area tube or channel about vii m long. Each aqueduct is individually cooled past pressurised water which is allowed to boil in the tube and emerges at about 290°C.

Refuelling

When fuel channels are isolated, the fuel assemblies can be lifted into and out of the reactor, allowing fuel replenishment while the reactor is in operation.

Graphite moderator

A series of graphite blocks environment, and hence separate, the pressure level tubes. They act as a moderator to slow down the neutrons released during fission so that a continuous fission concatenation reaction can exist maintained. Heat conduction between the blocks is enhanced by a mixture of helium and nitrogen gas.

Control rods

Boron carbide command rods absorb neutrons to control the rate of fission. A few brusque rods, inserted upwards from the bottom of the core, even the distribution of ability across the reactor. The primary control rods are inserted from the meridian downward and provide automatic, manual, or emergency command. The automated rods are regulated by feedback from in-core detectors. If at that place is a deviation from normal operating parameters (eastward.g. increased reactor ability level), the rods can exist dropped into the core to reduce or finish reactor activeness. A number of rods remain in the cadre during operation.

Coolant

Two separate water coolant loops each with 4 pumps circulate water through the pressure tubes to remove virtually of the heat from fission. At that place is also an emergency cadre cooling organisation which is designed to come into operation if either coolant circuit is interrupted.

Steam separator

Each of the two loops has ii steam drums, or separators, where steam from the heated coolant is fed to the turbine to produce electricity in the generator (each loop has a turbogenerator associated with it). The steam is then condensed and fed back into the circulating coolant.

Containment

There is no secure containment in the sense accepted in the Due west. The reactor core is located in a reinforced concrete lined cavity that acts as a radiation shield. The core sits on a heavy steel plate, with a 1000 tonne steel encompass plate on the peak. The extensions of the fuel channels penetrate the lower plate and the encompass plate and are welded to each. The steam separators of the coolant systems are housed in their ain concrete shields.

Positive void coefficient

The term 'positive void coefficient' is often associated with RBMK reactors.

Reactors cooled by boiling h2o will contain a sure corporeality of steam in the core. Because h2o is both a more than efficient coolant and a more than effective neutron absorber than steam, a change in the proportion of steam bubbles, or 'voids', in the coolant will event in a modify in core reactivity. The ratio of these changes is termed the void coefficient of reactivity. When the void coefficient is negative, an increment in steam will pb to a decrease in reactivity.

In those reactors where the same water circuit acts as both moderator and coolant, excess steam generation reduces the slowing of neutrons necessary to sustain the nuclear chain reaction. This leads to a reduction in power, and is a basic safety feature of most Western reactors.

In reactor designs where the moderator and coolant are of different materials, excess steam reduces the cooling of the reactor, only every bit the moderator remains intact the nuclear chain reaction continues. In some of these reactors, nearly notably the RBMK, the neutron absorbing backdrop of the cooling h2o are a significant factor in the operating characteristics. In such cases, the reduction in neutron absorption every bit a consequence of steam product, and the consequent presence of extra free neutrons, enhances the chain reaction. This leads to an increase in the reactivity of the system.

The void coefficient is merely one correspondent to the overall ability coefficient of reactivity, only in RBMK reactors it is the dominant component, reflecting a high degree of dependence of reactivity on the steam content of the core. At the time of the accident at Chernobyl, the void coefficient of reactivity was so positive that it overwhelmed the other components of the ability coefficient, and the power coefficient itself became positive. When the power began to increase, more steam was produced, which in plough led to an increase in ability. The boosted heat resulting from the increase in ability raised the temperature in the cooling excursion and more steam was produced. More steam means less cooling and less neutron absorption, resulting in a rapid increment in ability to around 100 times the reactor's rated chapters.

The value of the void coefficient is largely determined past the configuration of the reactor core. In RBMK reactors, an important factor affecting this is the operating reactivity margin.

Operating reactivity margin

Although the definition is non precise, the operating reactivity margin (ORM) is essentially the number of 'equivalent' command rods of nominal worth remaining in the reactor core. The operators at Chernobyl seemed to believe that prophylactic criteria would be met then long as the lower limit for the ORM of fifteen equivalent rods was adhered to, regardless of the bodily configuration of the core. The operators were not aware of the 'positive scram' effect where, post-obit a scram signal, the initial entry of the control rods really added reactivity to the lower region of the core (run across section below on Post accident changes to the RBMK).

The ORM could have an extreme effect on the void coefficient of reactivity, every bit was the case for the core configuration of Chernobyl iv in the run-upwardly to the blow. Unacceptably large void coefficients were prevented for initial cores by increasing fuel enrichment levels, with the backlog reactivity balanced past fixed absorbers. Nevertheless, with increasing fuel burn-up, these absorbers could be removed to maintain the fuel irradiation levels - shifting the void coefficient in the positive direction and increasing the sensitivity of the coefficient to the extent of insertion of the command and protection rods.

Post accident changes to the RBMK

After the blow at Chernobyl, several measures were taken to amend the safety of RBMK plants. All operating RBMK reactors in the former Soviet Marriage had the post-obit changes implemented to amend operating prophylactic:

• Reduction of the void coefficient of reactivity.
• Improvement of the response efficiency of the emergency protection arrangement.
• Introduction of adding programs to provide an indication of the value of the operating reactivity margin (ORM, i.e. the effective number of command rods remaining in the core) in the command room.
• Prevention of the emergency safety systems from being bypassed while the reactor is operating.
• In order to ensure adequate subcooling at the core inlet, the avoidance of modes of operation that crusade a reduction in the difference from nucleate boiling (DNB) ratio of the coolant at the reactor inlet.

Measures to reduce the void coefficient of reactivity were carried out past:

• The installation of 80-90 additional fixed absorbers in the core to inhibit operation at low power.
• Increasing the ORM from 26-30 rods (in steady land operational manner) to 43-48.
• An increment in fuel enrichment from ii% to ii.4%.

The increase in the number of fixed absorbers and the ORM reduced the value of the void coefficient of reactivity to +ß (where ß is the effective delayed neutron fraction). The additional absorbers require the use of college fuel enrichment to compensate for the increased neutron assimilation.

The efficiency and speed of the emergency protection organization was improved by implementing three independent retrofitting operations:

• Retrofitting of command rods with a design that does non give rise to water columns at the lesser of the channels.
• Scram (shut downwardly) rod insertion time cutting from xviii to 12 seconds.
• The installation of a fast-interim emergency protection (FAEP) system.

I of the most important mail service-accident changes to the RBMK was the retrofitting of the control rods. A graphite 'displacer' is fastened to each end of the length of absorber of each rod (except for 12 rods used in automated control). The lower displacer prevents coolant water from entering the space vacated as the rod is withdrawn, thus augmenting the reactivity worth of the rod. However, the dimensions of the rod and displacers were such that, with the rod fully withdrawn, the iv.v m displacer sat centrally within the fuelled region of the core with 1.25 m of h2o at either stop. On a scram signal, equally the rod falls, the water at the lower part of the channel is replaced by the lesser of the graphite displacer, thus initially calculation reactivity to the bottom part of the core. Following the Chernobyl accident, this 'positive scram' effect was mitigated by retrofitting the control rods then that, with the rods fully retracted, there would not be a region containing water at the bottom of the core.

The FAEP organisation was designed so that 24 emergency protection command rods would insert negative reactivity of at least 2ß in nether two.v seconds. Tests in 1987-'88 at the Ignalina and Saint petersburg plants (the first RBMKs to be fitted with the new FAEP system) confirmed these characteristics.

In addition to the to a higher place changes, several further modifications accept been implemented at RBMK plants. These measures consist of:

• Replacement of the fuel channels at all units (except Smolensk 3).
• Replacement of the grouping distribution headers and addition of check valves.
• Improvements to the emergency core cooling systems.
• Improvements of the reactor cavity over-pressure level protection systems.
• Replacement of the SKALA process computer.

Operating RBMK plants

At that place are currently viii operating RBMKs, all of which are in Russia. One more was under construction in Russia (Kursk 5), just it is now cancelled. All operating RBMKs began functioning between 1979 (Kursk 2) and 1990 (Smolensk three). In that location are three distinct generations of RBMK reactors having significant differences with respect to their safety design features:

• The four offset-generation units are Leningrad 1&2 (both now shutdown), and Kursk 1 (now shutdown) and ii. They were designed and brought online in the early-to-mid-1970s, before new standards on the pattern and construction of nuclear ability plants, the OPB-82 General Safety Provisions, were introduced in the Soviet Union in 1982.
• 2nd-generation RBMKs, brought online since the late 1970s and early on 1980s include Petrograd 3&4; Kursk 3&iv; Ignalina 1 (now shutdown); and Smolensk 1&2. Ignalina 2 (now shutdown) had rubber features across those of other 2d generation units. These units conform to the OPB-82 standards.
• After the Chernobyl accident, Soviet safe standards were revised over again (OPB-88). One RBMK (Smolensk iii) has been built to these third-generation standards. Additional pattern changes were beingness incorporated in the structure of Kursk 5 before it was cancelled.

In 2006, Rosatom said it was because operating lifetime extensions and uprating of its operating RBMK reactors. Following significant blueprint modifications fabricated after the Chernobyl blow, besides as extensive refurbishment including replacement of fuel channels, a 45-year operating lifetime is seen as realistic for the grand MWe-class units. In 2021, they provided about 25% of Russia's nuclear-generated electricity.

In the Table beneath, the 'operating until' dates are the scheduled shutdown for these plants, with xv-year lifetime extensions in some cases. Republic of lithuania, on the other hand, airtight Ignalina 1&2 early as a status for entry into the European Union.

Russian federation's long-term plans had before included the possibility of replacing the Leningrad units, at the end of their extended service life, by new MKER-1000 units. These are a modification of the RBMK design. The main differences are in the spacing of the graphite lattice in the core and the incorporation of passive condom systems.

RBMK reactors

Location	Unit	Start power	Unit of measurement net capacity (MWe)	Status
Republic of lithuania
Ignalina	i	1983	1185 (originally 1300)	Closed 12/2004
	ii	1987	1185 (originally 1300)	Closed 12/2009
Russia
Kursk	ane	1976	925	Closed 12/2021
	2	1979	925	Operating until 2024
	3	1984	925	Operating until 2029
	4	1986	925	Operating until 2031
	5	-		Structure cancelled
Leningrad	1	1973	925	Closed 12/2018
	2	1975	925	Closed 11/2020
	3	1979	925	Operating until 2025
	4	1981	925	Operating until 2026
Smolensk	1	1983	925	Operating until 2028
	ii	1985	925	Operating until 2030
	three	1990	925	Operating until 2050
Ukraine
Chernobyl	ane	1977	925	Closed 1996
	ii	1978	925	Airtight 1991
	3	1981	925	Closed 2000
	iv	1983	925	Reactor destroyed April 1986
	5	-	925	Structure cancelled
	6	-	925	Structure cancelled

Operational dates are subject to licence extension – see information page on Nuclear Power in Russia.

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Notes & references

General sources

INSAG-7, The Chernobyl Blow: Updating of INSAG-1, A written report by the International Nuclear Safe Informational Group, International Atomic Energy Bureau, Safety Serial No. 75-INSAG-7, 1992, (ISBN: 9201046928). Come across especially: Annex I of INSAG-seven, the Report by a Commission to the USSR Land Committee for the Supervision of Safe in Industry and Nuclear Power (SCSSINP); and Addendum Ii of INSAG-7, the Report past Working Group of USSR Experts, Causes and Circumstances of the Blow at Unit 4 of the Chernobyl Nuclear Power Plant and Measures to Amend the Safety of Plants with RBMK Reactors (Moscow, 1991)

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Source: https://www.world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/appendices/rbmk-reactors.aspx