VVER
| VVER | |
|---|---|
| Developer / Manufacturer: | OKB Gidropress |
| Developing country: |
 Soviet Union
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| Reactor data | |
| Reactor type: | Pressurized water reactor |
| Design type: | pressure vessel |
| Moderator: | light water |
| Cooling: | light water |
| Vapor bubble coefficient: | negative |
| Power classes in MW (gross): | 210, 365, 440, 1000, 1160, 1200, 1300, 1500 |
| Containment: | from 3rd generation available, as well as with the export versions WWER-440/311 and WWER-440/318 |
| Copies built: | 66 |
Under the designation WWER (water-water power reactor, Russian Водо-водяной энергетический реактор , scientific. Transliteration Vodo-vodjanoj ėnergetičeskij reactor , transkr. Wodo-vodyanoi energetitscheski reactor , ВВЭР ), certain types of pressurized water reactors of Soviet or Russian combined type. The term water-water stands for water-moderated and water-cooled. The component, which usually fuel is called, is in the case of VVER reactors fuel assembly or shortly cassette ( Russian кассета ) .
Generations
There are four generations of reactors. The first number indicates the specific type of reactor; this usually corresponds to the approximate electrical output of the power plant in megawatts. The second number is the version of the reactor or the project name. The first two prototypes of this type of reactor (VVER-210 and VVER-365) were used and researched at the Novovoronezh nuclear power plant . The VVER-210 was developed at the Kurchatov Institute , all others were developed by the state-owned Soviet, later Russian company OKB Gidropress .
| Generation VVER |
Inferior reactors |
More powerful reactors |
Nuclear power plant |
|---|---|---|---|
| 1st generation | WWER-210 WWER-365 WWER-440/179 WWER-440/230 WWER-440/270 |
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| 2nd generation | WWER-440/213 WWER-440/311 WWER-440/318 |
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| 3rd generation | WWER-640/407 WWER-640/470 or WPBER-600 |
VVER-1000/187 | |
| WWER-1000/302 | |||
| VVER-1000/320 | |||
| WWER-1000/338 | |||
| WWER-1000/392 | AES-91 | ||
| WWER-1000/392 | AES-92 | ||
| WWER-1000/466 | |||
| VVER-1160 | |||
| WWER-1200/491 | AES-2006 | ||
| WWER-1500/448 |
Physical-technical data
The physical and technical data of the VVER type reactors (except VVER-1200) are from the standard work Atomic Energy in Science and Industry by Andranik Petrosʹjanc (1906-2005), who was chairman of the State Committee of the USSR for the use of atomic energy from 1978 to 1986 , taken.
| parameter | VVER-210 | VVER-365 | VVER-440 | VVER-1000 | VVER-1200 |
|---|---|---|---|---|---|
| Electric power (MW) | 210 | 365 | 440 | 1,000 | 1,200 |
| Thermal power (MW) | 760 | 1,320 | 1,375 | 3,000 | 3,200 |
| Gross efficiency (%) | 27.6 | 27.6 | 31 | 33 | 37 |
| Steam pressure in front of the turbine (Mpa) | 2.9 | 2.9 | 4.4 | 6th | 7th |
| Primary circuit vapor pressure (Mpa) | 10 | 10.5 | 12.5 | 16 | 16.2 |
| Number of coolant circuits | 6th | 8th | 6th | 4th | 4th |
| Coolant throughput (m³ / h) | 36,500 | 49,500 | 39,000 | 76,000 | 85,600 |
| Primary circuit inlet temperature (° C) | 250 | 250 | 269 | 289 | 298.6 |
| Mean temperature increase (° C) | 19th | 25th | 31 | 35 | 31.1 |
| Active zone diameter (m) | 2.88 | 2.88 | 2.88 | 3.12 | k. A. |
| Active zone height (m) | 2.50 | 2.50 | 2.50 | 3.50 | k. A. |
| Number of fuel assemblies | 343 | 349 | 349 | 151/163 | 163 |
| Diameter of a fuel rod (mm) | 10.2 | 9.1 | 9.1 | 9.1 | 9.1 |
| Number of fuel rods per cartridge | 90 | 126 | 126 | 312/331 | 312 |
| Grid step size (mm) | 14.3 | 12.2 | 12.2 | 12.6 | k. A. |
| Number of rule cassettes | 37 | 73 | 37 | 109 | 121 |
| Uranium loading (t) | 38 | 40 | 42 | 66 | 76-85.5 |
| Mean uranium enrichment (%) | 2.0 | 3.0 | 3.5 | 4.26 | 4.69 |
| Burnup (MWd / kg) | 13 | 27 | 28.6 | 26-60 | to 70 |
| Expected service life (years) | 20th | 20th | 40 | 40-50 | 60 |
Some data from the newer reactor types may differ slightly depending on the source.
VVER-210
The prototype of all Soviet / Russian pressurized water reactors was the type VVER-210. It was developed under the project name W-1 at the Kurchatov Institute and built as the first block of the Novovoronesch nuclear power plant. The physical start "with the cover open" was carried out in December 1963, on September 8, 1964 the reactor went critical. On September 30th, it was connected to the grid and on December 27th, 1964 it reached its design stage. At that time it was one of the most powerful nuclear reactors in the world.
The following technical solutions in particular were tested in this reactor project:
- The hexagonal cassette shape,
- Materials for the fuel element cladding,
- Materials, shape, body and mounting of the reactor,
- Control systems and reactor safety and
- Temperature control and energy release .
In 1984 this first unit was taken out of service.
VVER-70
From January 1957, OKB Gidropress developed a variant of the pressurized water reactor WWER-210 with a gross electrical output of 70 MW for the planned Rheinsberg nuclear power plant under the project name W-2 . This type of reactor was named VVER-70. At the end of 1958, the technical design of the W-2 reactor was completed. It should be noted that the W-1 and W-2 projects were developed only a short time apart, so that many technical solutions were similar.
Construction work on the Rheinsberg nuclear power plant began on January 1, 1960. The reactor became critical for the first time on March 11, 1966 . The commissioning ceremony took place on May 9, 1966. The reactor reached its design stage on October 11, 1966. This also began continuous commercial operation.
This type of reactor was designed for an operating time of 20 years. In 1986, after renovation work, it was extended by five years, so regular shutdown was scheduled for 1992. Due to serious safety concerns, the nuclear power plant was taken out of service on June 1, 1990.
VVER-365
In the next stage, the VVER-365 reactor type with a thermal output of 1,320 MW was developed. The work on this was started after a government decree of August 30, 1962.
The most important innovations of the VVER-365 included:
- The mean temperature increase of the moderator, heat carrier and coolant water in the active zone was increased from 19 ° C to 25 ° C,
- two more circuits have been added to keep the dimensions of the main circulation pumps constant as the flow and pressure of the coolant increase,
- the principle of "dry" reloading of cassettes was adopted,
- burnable absorbers were used for the first time,
- a universal type of control cassette was developed and
- the non-uniformity of the neutron flux in the reactor core has been reduced.
In addition, the sum of all the surfaces of the fuel rods has been increased by reducing their diameter from 10.2 to 9.1 mm. At the same time the type of cassette was changed. The number of fuel rods per cartridge has increased from 90 to 126 fuel rods. This in turn resulted in a number of other design changes, both with regard to the geometry and the manufacture of cassettes and fuel rods as well as the reactor core itself.
The VVER-365 was built as the second block of the Novovoronezh nuclear power plant and commissioned in 1969. The reactor reached its design performance in April 1970. In 1990 the VVER-365 was taken out of service as planned.
VVER-440
![WWER-440: Sectional graphic of a fuel cell made of water (blue) with an addition of boric acid, a shell made of zirconium alloy (green), the fuel rod made of sintered uranium dioxide (orange) and a central cavity. Helium is injected between the fuel rod and the zircalloy tube (yellow). The outer diameter of a fuel rod is 7.6 mm, that of the zircalloy tube 9.1 mm [8]](https://de.wikipedia.org/wiki/Datei:VVER_Cell01_RK01.svg)
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WWER-440: Sectional graphic of a fuel cell made of water (blue) with an addition of boric acid , a shell made of zirconium alloy (green), the fuel rod made of sintered uranium dioxide (orange) and a central cavity. Helium is injected between the fuel rod and the zircalloy tube (yellow). The outer diameter of a fuel rod is 7.6 mm, that of the zircalloy tube 9.1 mm
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Cell for detectors made of water (blue) with an addition of boric acid and a guide tube made of zircalloy (green)
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The series WWER-440 includes the old type WWER-440/230 and the newer type WWER-440/213, which has been improved in essential areas. There is also a special type that was developed only for the Finnish nuclear power plant Loviisa in order to meet the safety requirements there. Like all pressurized water reactors, the WWER-440 uses water both to cool the reactor core and to generate steam as well as to moderate the neutrons. Low-enriched uranium dioxide is used as fuel. One of the special features of the WWER-440/230 is the erection of double blocks with a common machine house .
According to the manufacturer, the radioactive dose rate in the vicinity of a nuclear power plant of the type VVER-440 increases by less than 0.5 mSv per year.
For the transport and the intermediate storage of the fuel elements, for example, Castor casings from GNS can be used, which were specially developed for the VVER-440 series. The type CASTOR 440/84 can hold 84 fuel elements. It is 4.08 m long and has a diameter of 2.66 m. Its mass is 116 tons.
The WWER-440 has a particularly slim reactor pressure vessel. The reactor core is therefore close to the steel walls, the water-filled gap between them is only sixteen centimeters wide, which is much narrower than in most nuclear power plants built in the west. The neutrons are slowed down less strongly in this narrow gap, so that the radiation exposure of the steel is higher and it therefore ages faster or becomes brittle .
An EU-funded research project called "Long Life" researched the embrittlement processes of various steel alloys under the influence of neutrons from 2010 to 2014. It was coordinated by scientists from the Helmholtz Center Dresden-Rossendorf under the direction of Eberhard Altstadt . The Helmholtz Center also examined steel samples from three blocks of the Greifswald nuclear power plant of the VVER type, which was operated from 1973 to 1990 . Due to the different operating times of the blocks, the steel used in them was irradiated with neutrons to different degrees. The embrittlement of the steel can thus be determined depending on the neutron bombardment and compared with the previous guide values for the aging of steel in nuclear power plants.
WWER-440/230
The reactors of the first VVER generation 230 have a number of safety deficiencies:
- low redundancy of the safety devices
- no all-enclosing security container
- Inadequate emergency cooling if a main coolant line breaks
- poor spatial separation of the (redundant) safety devices
- Confusing and outdated control technology and control fittings
VVER-440/230 series reactors were in operation in Kozloduj and Bohunice , among others . The European Union had declared that VVER-440/230 reactors “cannot be upgraded to the required level of safety” and will therefore have to be decommissioned when the relevant countries join the EU - the corresponding VVER-440 / 230s were decommissioned by 2007 . In the GDR this type of reactor was in use in Greifswald and - like all other nuclear power plants in the GDR - was decommissioned in the course of reunification.
WWER-440/213
With the type WWER440 / 213 numerous defects have been corrected. The emergency cooling system is now able to intervene effectively in the event of any defects in the coolant supplier. In addition, the safety systems were designed with triple redundancy and fire protection was significantly improved. This series also has an attached bubble condenser . This gives the radioactive vapor released by a leak - even a large one - more space and can also condense in water reservoirs before the design pressure is reached.
In addition to WWER-440/230, a reactor of the type WWER-440/213 was also in operation in Greifswald - this was also shut down after 1989. Three more were under construction, but never went online. VVER-440/213 series reactors are located in the EU in Dukovany , Bohunice , Mochovce and Paks .
WWER-440/318
An export version of the WWER-440/213 is the WWER-440/318. It was to be used in the Juraguá nuclear power plant . In contrast to the standard series 213, the WWER-440/318 has a containment .
VVER-1000
The WWER-1000 is a further development of the WWER-440 with improved safety devices - including a safety container - and higher electrical power (1,000 MW), with proven components being taken over from the WWER-440. The VVER-1000 reactors can be upgraded to a higher level of safety with a corresponding effort. The entire control technology as well as the slow computers have to be replaced. Furthermore, part of the still user-unfriendly monitoring systems and displays are being modernized. The WWER-1000 uses cooling pumps of the type GCNA-1391 with an internal requirement of 5 MW per pump. The pump speed is 1000 revolutions per minute. The steam generator of the VVER-1000 is of the type ПГВ-1000М.
VVER-1000/320 series reactors are located in Balakowo (Russia), Kalinin (Russia), Kozloduy (Bulgaria), Temelín (Czech Republic), Khmelnyzkyj (Ukraine), Rvine-3 and Rivne-4 (Ukraine) and Zaporizhia (Ukraine).
The reactors of the VVER-1000/392 found in nuclear power plants designated AES-91 , and AES-92 used (see Atomstroiexport ) . The first nuclear power plant of the type AES-91 was built in Tianwan (People's Republic of China) with a reactor VVER-1000/428 adapted for this project. The version adapted for India is called WWER-1000/412 and is used in the Kudankulam nuclear power plant of the AES-92 type. Both have been equipped with Western control systems; more passive safety devices were provided for the AES-92 variant. In contrast to the AES-92 type, the AES-91 nuclear power plant has additional protection against earthquakes.
According to the manufacturer, the outbreak of corium (mixture of fuel and material of the fuel rod cladding) after a core meltdown is impossible with VVER from the series VVER-1000/320 . For this purpose, the reactor pressure vessel is cooled from the outside by passive measures so that the steel of the reactor pressure vessel still has sufficient strength to keep the melt inside. Since the research into core meltdown is only at the basic scientific stage, no guarantee can be given regarding the controllability of core meltdown scenarios.
For some time now, experiments have been carried out with new types of fuel assemblies for all VVER reactors. The plan is to recycle the spent fuel elements from the RBMK reactors and use them as fuel elements for VVER reactors. These are up to 2.5% more efficient than conventional VVER fuel elements. The fuel is currently in experimental use in the reactors of the Kalinin nuclear power plant . The spent fuel elements can in turn be processed into MOX fuel elements , which have been used in the Belojarsk nuclear power plant since the beginning of 2008 .
According to the manufacturer, the radioactive dose rate in the vicinity of a nuclear power plant of the type VVER-1000 increases by less than 0.5 mSv per year.
VVER-1200
The WWER-1200 reactor is a further development of the WWER-1000 reactor and the AES-91 and AES-92. The basis for the development of the reactor was the construction of the Tianwan nuclear power plant and the Kudankulam nuclear power plant. The WWER-1200/491 was then developed from their technology and safety systems and an increase in performance was achieved. This type of reactor is to be used in a newly designed nuclear power plant AES-2006 , a generation III + reactor . The reactor was developed by OKB Gidropress in cooperation with Atomstroiexport , a company founded in 1998 . The first reactors in Novovoronezh and Leningrad have already been completed. The WWER-1200 reactor is designed for a service life of 60 years. What will be new about these VVERs is the high-speed steam turbine, which is only used in new types of nuclear reactors. As with the WWER-1000, pumps of type GCNA-1391 and steam generators of type PGV-1000 MKP are also used in WWER-1200.
Differences between the VVER-1200 and the VVER-1000 are, for example:
- larger diameter of the reactor vessel
- more efficient use of fuel rods
- possible increase in the thermal reactor output from 3200 MW to 3300 MW
| parameter | VVER-1200 |
|---|---|
| Length of reactor pressure vessel (m) | 11.185 |
| Reactor pressure vessel diameter (m) | 4,250 |
| Mass pressure vessel (t) | 330 |
| Diameter of the steam generator (m) | 4.2 |
| Total volume of pressurizer (m³) | 79 |
| Water volume pressurizer (m³) | 55 |
| Nominal pressure pressurizer outlet (MPa) | 16.1 |
| Pressurizer temperature (° C) | 347.9 |
| Utilization rate (%) | 90 |
| Cost per kW ($) | 2100 |
| Construction time (months) | 54 |
In the course of the 2007-2015 project , a plan was drawn up to meet Russia's growing energy needs and to take the old reactors off the grid. Among other things, the VVER-1200 (AES-2006) was used. A total of 28 reactors are planned. The first reactors will be built in the Novovoronezh II nuclear power plant . A VVER-1160, which is being built in Leningrad II , is to be built on the basis of the VVER-1200.
See also
Web links
Individual evidence
- ↑ Export version of the WWER-440/213
- ↑ Андраник Мелконович Петросьянц: Атомная энергия в науке и промышленности . Энергоатомиздат, Москва 1984, p. 158 (447 pp., Biblioatom.ru ).
- ↑ Нововоронежская АЭС-2. (PDF) Проект «АЭС-2006». Атомэнергопроект, accessed May 24, 2020 .
- ↑ М. П. Никитенко: РЕАКТОРНЫЕ УСТАНОВКИ ВВЭР. (PDF) ОКБ «Гидропесс», October 22, 2013, accessed on May 24, 2020 .
- ↑ Андраник Мелконович Петросьянц: Атомная энергия в науке и промышленности . Энергоатомиздат, Москва 1984, p. 143 (447 pp., Biblioatom.ru ).
- ↑ Реакторная установи ВВЭР-365 (В-ЗМ). Retrieved May 25, 2020 .
- ↑ Нововоронежская АЭС. Общая характеристика НВАЭС. Retrieved May 25, 2020 .
- ↑ a b c Б. А. Дементьев: Ядерные энергетические реакторы . Энергоатомиздат, Москва 1984, p. 18-21, 257 (280 pp.).
- ↑ a b Rosenergoatom - Radiation safety of the population and the environment - data on emissions ( Memento from February 28, 2014 in the Internet Archive ) (English)
- ↑ Hans-Joachim Elwenspoek: Imagine the CASTOR is coming ... Press and information office of the German Atomic Forum, Berlin 2006.
- ↑ Uta Bilow: Reactors under constant fire in: FAZ from September 22, 2010
- ^ H. Karwat: The evaluation of the bubble condenser containment of VVER-440/213 plants . Ed .: Technical University of Munich, Chair for Reactor Dynamics and Reactor Safety. December 22, 1999, doi : 10.1016 / 0029-5493 (95) 01062-M .
- ↑ NEI Source Book: Fourth Edition (NEISB_3.2) ( Memento from March 30, 2008 in the Internet Archive ) (English)
- ↑ NTI - Russia, Cuba, and the Juragua Nuclear Plant (English)
- ↑ a b World Nuclear Association: Nuclear Power in Russia (English)
- ↑ a b VG Asmolov et al. : New generation first-of-the kind unit - VVER-1200 design features . In: Nuclear Energy and Technology . tape 3 , no. 4 , 2017, p. 260-269 , doi : 10.1016 / j.nucet.2017.10.003 ( online ).
- ↑ Details about the VVER ( Memento from September 28, 2007 in the Internet Archive ) (English)