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Skip Navigation LinksNuclear Power Plant Types

 
Nuclear power plants are thermal power plants that obtain heat not by burning certain fuel, but by a nuclear chain reaction occurring in their reactors, through the fission or fusion of atoms.
 

In technical language, the union of nuclei is called fusion; however, its industrial utilisation has not yet been solved. On the other hand, fission reactors, i.e. those working on the principle of splitting nuclei with large mass numbers, have been serving humanity for more than half a century.

 

Fission reactor types:
 

1. Thermal reactor: nuclear fission is triggered and sustained by neutrons slowed down from a high energy level to thermal speed (see Information on Nuclear Energy / Nuclear Fission and Chain Reaction for more detail). Thus a neutron-slowing moderator is used for maintaining the chain reaction. Its types are:

 1.1. Light Water Reactors
      1.1.1.  Boiling Water Reactor
      1.1.2.  Pressurised Water Reactor
 1.2. Heavy Water Reactors
      1.2.1.  CANDU Reactor
 1.3. Graphite-moderated Reactors
      1.3.1.  Gas-cooled Reactor
      1.3.2.  RBMK Reactor 

2. Fast reactor: The chain reaction is not based on thermal neutrons, there is no moderator, fission is performed by fast neutrons.

 

Natural nuclear reactors

The world’s first nuclear reactors operated in West Africa, in the territory of today’s Gabon. Nearly two billion years ago the uranium isotope with mass number 235 was still present in natural uranium at that location in such proportions (3.7% compared to the present 0.7%, which corresponds to the level of enrichment of fuel used in today’s nuclear power plants) that under appropriate conditions (in the presence of water in uranium-containing rocks) a chain reaction could take place. These reactors could reach a thermal output of up to 20 kW, and the water in the rocks served as moderator material required for triggering a chain reaction. Upon the start of the chain reaction, the water warmed up, then reached its boiling point. Since in its steam phase, water could no longer perform its task of slowing down neutrons, the chain reaction stopped in the absence of moderation. As the chain reaction stopped, the temperature of the natural nuclear pile also began to decrease, then the steam was condensed, the water so produced functioned again as a moderator and the chain reaction started again. This process could repeat itself until the right conditions existed (water was present) or until the U235 content of uranium ore decreased to a level that it could no longer maintain a natural chain reaction.

 

1.1 Light Water Reactors

According to the working principle of Light Water Reactors, the tasks of both the moderator and heat transfer are performed by the cooling water. The coolant is boron-containing, treated natural water.

 

1.1.1 Boiling Water Reactor

This is one of the oldest types which operates on the simplest principles. The thermal energy released in the reactor during the nuclear chain reaction boils the feedwater and the generated steam is conveyed to a turbine. After condensation, the water is returned to the reactor by means of a feedwater pump after preheating.

 
Figure 1: Scheme of boiling water reactor
(Source: commons.wikimedia.org)
 
1. Reactor pressure vessel ​10. Generator
​2. Fuel assembly​ ​11. Exciter
​3. Control rod ​12. Condenser
​4. Circulating pump ​13. Cooling water
​5. Control rod drive ​14. Feedwater preheater
​6. Fresh steam ​15. Feedwater pump
​7. Feedwater ​16. Cooling water pump
​8. Steam turbine high pressure casing ​17. Concrete radiation protection
​9. Steam turbine low pressure casing ​18. To transmission line
 
 

1.1.2 Pressurised Water Reactor

As opposed to Boiling Water Reactors, it involves a system with two cooling circuits separated from each other, where the so-called primary coolant water is heated up in the reactor, then, after being conveyed to a steam generator, transfers its thermal energy to the secondary coolant, and afterwards, after cooling down, it returns to the reactor. The primary coolant is kept under high pressure, thus it does not reach its boiling point even at a temperature of about 300°C. The secondary coolant has a lower pressure, thus when heat energy is transferred to it, it not only warms up, but also reaches its boiling point and the generated steam, after entering the turbine, drives the turbine blades. In the condenser, the steam is condensed and, after being preheated, returns to the steam generator. 

The advantage of this design over the boiling water type with a single cooling circuit is that there is also a closed water circuit between the primary coolant entering the reactor and the external water source used for condensation. Thus the two media cannot mix with each other even in the case of an occurring leakage. This technological solution further decreases the chance of radioactively contaminated medium being released into the environment.

 
​Figure 2: Scheme of pressurised water reactor
(Source: wikipedia.org)
 
1. Reactor pressure vessel ​11. Exciter
​2. Fuel assembly    ​12. Condenser
​3. Control rod ​13. Cooling water
​4. Control rod drive ​14. Feedwater preheater
​5. Pressuriser ​15. Feedwater pump
​6. Steam generator ​16. Cooling water pump
​7. Feedwater ​17. Circulating pump​
​8. High pressure steam turbine ​18. To transmission line
​9. Low pressure steam turbine​
10. Generator 
​19. Fresh steam
20. Concrete radiation protection, containment​
 
 

1.2 Pressurised Heavy Water Reactor

This type was developed in Canada under the name of CANDU in the 1950s. Similarly to the pressurised water type, it has two cooling circuits. However, heavy water in the primary circuit is used only for heat transfer. Heavy water (the hydrogen atoms contains one neutron and one proton in the heavy water. This hydrogen isotope is known as deuterium) is also used for moderation, which allows natural uranium to be used as fuel. Due to the design of the CANDU reactors, it is possible to refuel the reactor even during operation. At the same time, the disadvantage of this technology is the high production cost of heavy water.

 
Figure 3: Scheme of pressurised heavy water reactor
 

1.3 Graphite-moderated Reactors

1.3.1 Gas-cooled Reactor

This reactor type was developed in the United Kingdom. It also contains two cooling circuits, uses graphite as moderator and carbon dioxide as coolant. Thus, a much higher primary temperature can be attained than with a water coolant. High temperature carbon dioxide evaporates the water in the secondary circuit through a heat exchanger.

 ​Figure 4: Scheme of the gas-cooled reactor
(Source: commons.wikimedia.org) 

1.     Refuelling tubes
2.     Control rods
3.     Graphite moderator
4.     Fuel assembly
5.     Reactor pressure vessel
6.     Gas circulating pump
7.     Feedwater
8.     Feedwater pump
9.     Heat exchanger
10.  Steam

 

1.3.2 RBMK Reactor (Light Water Cooled Graphite-moderated Reactor)

This is a reactor developed in the Soviet Union, which is also suitable for producing plutonium in addition to energy production. Slightly enriched or natural uranium is used as its fuel. The neutrons are slowed down by graphite moderator. Similarly to Boiling Water Reactors, water reaches its boiling point in the reactor and is conveyed to a turbine. The great disadvantage of this type is that moderation does not stop if the coolant is lost, thus the chain reaction does not stop either, as opposed to the water-moderated solutions. The Chernobyl Nuclear Power Plant, which suffered a tragic accident, was of such a type. Reactors of this type are no longer built today, because of the safety deficiencies.

 
Figure 5: Scheme of an RBMK reactor
 

2. Fast Reactors

Fast reactors were named after the fast neutrons determining the neutron spectrum of their core (i.e. the energy distribution of neutrons). There is no moderator in the reactor, so the neutrons do not slow down, i.e. they have much higher average energy than in thermal reactors. At such neutron energy level, the neutrons interact with the different isotopes in a different way than in the thermal spectrum. The propensity for fission of isotopes that cannot be splitted by slow neutrons slightly increases, but it is even more important that the number of neutrons produced by fission increases. Absorbed by non-fissile heavy nuclei (fertile isotopes), the additional neutrons are able to convert them into fissile ones, this phenomenon is called breeding.

The type of fast reactor that is able to produce more fissile materials by the end of the fuel campaign than the amount of fissile materials in the core is called a breeder reactor. In these reactors, isotope 238 accounting for the majority of uranium is converted into the fissile isotope 239 of plutonium. However, in order for this process to be able to occur and for a positive fissile material balance, a much higher initial fissile material content is required compared to thermal reactors: it may be enriched to 20%.

Not only uranium, but also thorium can be used as fertile material, which is converted into fissile uranium-233 isotope after neutron capture. At present, the thorium fuel cycle is not at the same technological level as the uranium-plutonium one; however, because of its numerous advantages, intensive research and development are underway in this area, too. Another type of fast reactor is the one that splits heavy nuclei filled into its core (mainly plutonium and heavy transuranium elements), but does not accumulate new fissile material. Such reactors are called burners. They are used for the conversion and destruction of hazardous radioactive wastes with long half-lives.

The coolant of fast reactors can be only a material that does not moderate neutrons, thus water cooling is excluded in this case. The most widespread is the liquid metal, i.e. sodium cooling. In Russia, sodium-cooled fast reactors are operational and being constructed for commercial purposes even at present. In this case, the circulating pumps circulate liquid sodium through the core instead of water. The other two widespread fast reactor coolants are liquid heavy metals: liquid lead or a mixture of lead and bismuth (eutectic) and gas cooling: mostly helium and carbon dioxide.

 
Figure 6: Scheme of a sodium-cooled fast reactor
(Source: npp.hu)
 
 
​​​1 Fuel (fissile material) 
​​9 Lid        
​17 Condenser
​2 Fuel (breeding material)
​10 Na/Na heat exchanger
​18 Cooling water
​3 Control rods (boron carbide)
11 Secondary Na   
​19 Cooling water pump
​4 Primary Na pump 
​​​​12 Secondary Na pump
​20 High pressure turbine
​5 Primary Na   
​​13 Steam generator
​21 Low pressure turbine
​6 Reactor vessel
14 Fresh steam 
​22 Generator
​7 Protective vessel 
​15 Feedwater preheater 
​23 Reactor building
​​8 Reactor vessel head ​​
​​16 Feedwater pump

 

Paks Nuclear Power Plant

Unit 1 of the Paks Nuclear Power Plant started its operation in 1982. Three additional units followed it in 1984, 1986 and 1987. The original gross electrical power of the Generation II VVER-440 water-moderated and water-cooled units, which belong to the pressurised water reactor family and were built with Russian technology, was 440 MW. During operation, this value was increased to 500 MW by increasing the power of the reactor and developing the turbine. The house load of the power plant is about 30 MW per unit, thus the electricity fed to the grid today reaches 1,880 MW.

The Paks units operate with UO2-containing fuel enriched with the U235 isotope, which is placed in fuel rods. In order to facilitate fuel handling, these rods are arranged in fuel assemblies. Each fuel assembly contains 126 fuel rods, and 312 of such fuel assemblies are placed in the reactor together with 37 control and safety assemblies. The safety assemblies are responsible for the immediate shutdown of the reactor in the case of an emergency shutdown, and the control rods control the output of the reactor. In an operational reactor, the thermal energy produced by nuclear chain reaction is conveyed to a steam generator by primary circuit water circulated by a high performance pump, where it transfers it to the secondary circuit through heat-exchanger tubes. At high pressure (123 bar), the primary coolant with a temperature of 300°C does not reach its boiling point. The primary coolant enters the steam generator at such a temperature, then it is returned to the reactor after cooling by about 30°C, where it is reheated. There are six such loops per unit, i.e. six main circulating pumps and steam generators belong to one reactor.
 

There is a so-called secondary coolant on the other side of the heat-exchanger tubes. The medium enters the steam generator at about 220°C; it is not only heated, but also reaches its boiling point. Saturated steam with a pressure of 46 bar and a temperature of 246°C leaves the steam generator and is conveyed to the turbine. The steam drives the turbine blades and the rotational energy so generated is converted into electricity by a generator.

The waste steam leaving the turbine is caused to condense with cooling water abstracted from the Danube through a heat exchanger, the so-called condenser. The condensed water is preheated, then is returned to the steam generator using a feedwater pump, while the cooling water is returned to the Danube with a temperature about 8°C higher.

The originally planned 30 years operating lifetime of the units would expire in the current years, between 2012 and 2017. However, due to continuous development and major safety-enhancing modifications, they are in an outstanding technical condition, which allows their lifetime to be extended by an additional 20 years. Unit 1 and 2 have already been granted a licence from the Hungarian Atomic Energy Authority for the extension of its lifetime by 20 years, and the licensing of the lifetime extension of the other units is currently underway.

 

Generation III nuclear power plants 

The units being constructed today belong to Generation III of nuclear power plants, which have significant technical, safety and economic advantages over the previous types:

  • reduction of construction costs and time due to standard designs, previously every unit meant a new power plant design;
  •  longer operating life: 50 to 60 years;
  • higher availability: up to 18 to 24-month fuel cycles, shorter reloading time and a higher load factor;
  • standardisation, passive protection systems: easier and safer operation;
  • management and prevention of severe accidents;
  • more optimal fuel use: better efficiency of use, less radioactive waste;
  •  in the case of an accident, the unit is able to shut down without operator action and to operate its protection system without an external power source for up to 72 hours;
  • protection against aeroplane crash, earthquakes and other environmental disasters: double wall protective building, a so-called containment;
  • in the case of an accident resulting in a core meltdown, a protection system preventing the release of the corium into the environment, e.g. a core catcher.
 

Sources used

http://npp.hu​

http://www.world-nuclear.org

http://wikipedia.org