VVER
This article is licensed under the GNU Free Documentation License. It uses material from the Wikipedia article "VVER".

WWER-10ff (or VVER-1000 as a direct transliteration of Russian ВВЭР-1000) is a 1000MW Russian nuclear power reactor of PWR type.

The VVER (From Russian: Водо-водяной энергетический реактор; transliterates as Vodo-Vodyanoi Energetichesky Reactor; Water-Water Energetic Reactor) is a series of pressurised water reactors (PWRs) developed by the former Soviet Union (FSU) and used by FSU satellites, China, Finland and the present-day Russian Federation.

The earliest VVERs were developed before 1970. The VVER-440 Model V230 is the most common design, delivering 440MW of electrical power. The V230 employs six primary coolant loops each with a horizontal steam generator. A modified version of VVER-440, Model V213, was a product of the first nuclear safety standards adopted by Soviet designers. This model includes added emergency core cooling and auxiliary feedwater systems as well as upgraded accident localization systems. The larger VVER-1000 was developed after 1975 and is a four-loop system housed in a containment-type structure with a spray steam suppression system. VVER reactor designs have been developed to incorporate automatic control, passive safety and containment systems associated with Western third generation nuclear reactors.

VVER series nuclear reactors were also scaled down in size and used by the Soviet Navy and RFAS nuclear submarine fleet as well as by surface warships.

The Russian abbreviation VVER stands for water-cooled, water-moderated energy reactor. This describes the pressurized water reactor design. Reactor fuel rods are fully immersed in water kept at 15MPa of pressure so that it does not boil at normal (220 to over 300°C) operating temperatures. Water in the reactor serves both as a coolant and a moderator which is an important safety feature. Should coolant circulation fail the neutron moderation effect of the water diminishes, reducing reaction intensity and compensating for loss of cooling, a condition known as negative void coefficient. The whole reactor is encased in a massive steel pressure shell. Fuel is low enriched (ca. 2.4–4.4% ²³⁵U) uranium dioxide (UO2) or equivalent pressed into pellets and assembled into fuel rods.

Intensity of the nuclear reaction is controlled by control rods that can be inserted into the reactor from above. These rods are made from a neutron absorbing material and depending on depth of insertion hinder the chain reaction. If there is an emergency, a reactor shutdown can be performed by full insertion of the control rods into the core.

Primary cooling circuit

As stated above, water in the primary circuit is kept under constant pressure to avoid boiling. Since the water transfers all the heat from the core and is irradiated, integrity of this circuit is most crucial. In the circuit four subsystems can be distinguished:

1. Reactor: Water flows through fuel rod assemblies and is heated by the nuclear chain reaction.
2. Pressurizer: To keep the water under constant but controlled pressure, the pressurizer regulates pressure by means of electrical heating and relief valves.
3. Steam Generator: In the steam generator, heat from primary coolant water is used to boil water in the secondary circuit.
4. Pump: The pump ensures proper circulation of the water through the circuit.

To ensure safety primary components are redundant.

Secondary circuit and electrical output

The secondary circuit also consists of different subsystems:

1. Steam Generator: Secondary water is boiled taking heat from the primary circuit. Before entering the turbine remaining water is separated from the steam so that the steam is dry.
2. Turbine: The expanding steam drives a turbine, which connects to an electrical generator. The turbine is split into high and low pressure sections. To prevent condensation (Water droplets at high speed damage the turbine blades) steam is reheated between these sections. Reactors of the VVER-1000 type deliver 1GW of electrical power.
3. Condenser: The steam is cooled and allowed to condense, shedding waste heat into a cooling circuit.
4. Deaerator: Removes gases from the coolant.
5. Pump: The circulation pumps are each driven by their own small steam turbine.

To increase efficiency of the process, steam from the turbine is taken to reheat coolant before the deaerator and the steam generator. Water in this circuit is not supposed to be radioactive.

Cooling circuit

The cooling circuit is an open circuit diverting water from an outside reservoir such as a lake or river. Evaporative cooling towers, cooling basins or ponds exhaust waste heat from the generation circuit, releasing it into the environment. In addition to generating electricity most VVERs have a capability to supply heat for residential and industrial use. Operational examples of such systems are the plants at Bohunice and Dukovany. ¹

Safety barriers

The two VVER440 units in Loviisa have containment buildings that fulfil Western safety standards.

A typical design feature of nuclear reactors is layered safety barriers preventing escape of radioactive material. VVER reactors have four layers:

1. Fuel pellets: Radioactive elements are retained within the crystal structure of the fuel pellets.
2. Fuel rods: The zircaloy tubes provide a further barrier resistant to heat and high pressure.
3. Reactor Shell: A massive steel shell encases the whole fuel assembly hermetically.
4. Reactor Building: A concrete containment building that encases the whole first circuit is strong enough to resist the pressure surge a breach in the first circuit would cause.

Currently operating Russian VVERs are inherently safer designs than the RBMK reactors of Chernobyl infamy. The former Soviet Union opted to construct graphite-moderated RBMK series nuclear reactors without contaiment structures on grounds of cost as well as the relative ease of re-fueling RBMK reactors. An RBMK reactor can be re-fueled while still operational compared to the VVER which needs to be shut down. Many levels of protection and containment have both been proposed and constructed for RBMK and VVER type reactors, including safety systems meeting Western nuclear standards.

Operational life of VVER 1000

Reactorhall of the AES-92

When first built the VVER design was intended to be operational for 35 years. A mid-life major overhaul including a complete replacement of critical parts such as fuel and control rod channels was thought necessary after that.²Since RBMK reactors specified a major replacement programme at 35 years designers originally decided this needed to happen in the VVER type as well, although they are of more robust design than the RBMK type. Most of Russia's VVER plants are now reaching and passing the 35 year mark. More recent design studies have allowed for an extension of lifetime up to 50 years with replacement of equipment. New VVERs will be nameplated with the extended lifetime.

VVER-1200

The VVER-1200 is an evolution of the VVER-1000 being offered for export. Specifications include a $1,200/kW-electric capital cost, 54 month planned construction time, and expected 50 year lifetime at 90% capacity factor. The VVER 1200 will produce 1,200 MWe of power. Safety features include a containment building and missile shield. It will have full emergency systems that include an emergency core cooling system, emergency backup diesel power supply, advanced refueling machine, computerized reactor control systems, backup feedwater supply and reactor SCRAM system. The nuclear reactor and associated systems will be hosted in one single building and there will be another building for the turbogenerators. The main building will comprise the reactor, refueling machine and diesel backup power supply, steam generators and reactor control systems.

If a VVER 1200 experiences a loss of coolant accident or loss of power accident the turbogenerators 'coast down' for 30 seconds, during which a shutdown can be initiated using residual power in the system. Further emergency power is available from a backup set of diesel generators kept on standby to maintain cooling flow to the reactor. The reactor design has been refined to optimize fuel efficiency.

The first two units are proposed for Leningrad Nuclear Power Plant II and Novovoronezh Nuclear Power Plant II.

Power plants

See the Wikipedia pages for each facility for sources.

Russia recently installed two nuclear reactors in China at the Tianwan Nuclear Power Plant, and an extension consisting of a further two reactors is now at the contracting stage. This is the first time the two countries have co-operated on a reactor project. The reactors are the VVER 1000 type, which Russia has improved incrementally while retaining the basic design. These VVER 1000 reactors are housed in a confinement shell capable of being hit by an aircraft weighing 20 tonnes and suffering no expected damage. Russia delivered initial fuel loads for the reactors but China plans to begin indigenous fuel fabrication for the Tianwan plant in 2009.³ The IAEA has referred to the station as the "safest nuclear power plant in the world".⁴

The improved VVER 1000 uses many third party parts to form the reactor. While the main reactor and turbogenerators are of Russian design, the control room is designed and built by an international consortium. In this way new VVER 1000 plants meet widely recognised safety standards. These include the emergency core cooling system and confinement system. Safety systems were already mostly in place but previous monitoring of these systems did not meet international safety standards. The new VVER 1000 plant built in China has 94% of its systems automated, meaning the plant can control itself under most situations. Even refueling procedures require little human intervention. Five people are still needed in the control room.

References

External links

• THE KUDANKULAM ATOMIC POWER PROJECT
• Reactor VVER-1000, Rosenergoatom
• VVER Reactor, Virtual Nuclear Tourist
• VVER Design, International Nuclear Safety Center, Argonne National Laboratory
• The VVER-440/213 reactor type
• VVER Reactors, at the US Energy Information Administration site.