Sections: Introduction | Process | Components | Classification
Subsections: Gas-Cooled | Light-Water | Heavy-Water | Fast Breeder | PRISM

Nuclear Power Plants

Introduction Development in nuclear power for the future is following three main paths: the extension of present reactor designs to create advanced concepts with passive safety features and improved economics; the development of breeder reactors to extend supply of fissionable fuel; and the development of a new nuclear energy source, fusion. Thus, classification of present reactors might prove to be preemptive due to fast development in research. It is, however, imperative that a classification be done, not to stereotype each, but to make a general description of the similar reactor technologies available at present. An attempt is also done to include those that are presently called experimental stage, of which mostly are either improvement in previous design or technology. Note: The following classification of Nuclear Fission Power Plants is made only for the purpose of learning the different reactor technology present at this time. It may not be a comprehensive list. For new development refer to any technical magazine on Nuclear Power engineering or on the internet.

Classification Nuclear Fission Power Plants can be classified in a number of ways. Below is a classification made by this author, culled from readings in Nuclear Engineering and current magazine issues of Asian Electricity, Modern Power Systems, Refocus, and Power Engineering. [Check the list of Related Materials on Nuclear Power Plant Engineering]

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Subsections: Gas-Cooled | Light-Water | Heavy-Water | Fast Breeder | PRISM

Gas-Cooled Reactor [GCR]

Moderator: Carbon
Coolant: Helium or Carbon Dioxide gas
Fuel: hexagonal graphite blocks or rods of uranium carbide, UC₂, fissile material coated with pyrolytic-carbon. For fertile material, Thorium oxide [ThO₂] is also used where: ²³²Th is converted to ²³³U. The fertile material is also coated with two layers of pyrolitic carbon. The two type of fuel particles are mixed in the proper proportions and are formed with a carbon matrix into fuel rods 15.6m diameter, 60 mm long. These rods are inserted into the holes in the graphite blocks with Helium coolant flow.

Magnox Gas-Cooled Reactor [MGR], uses graphite as a moderator. Fuel is usually natural (unenriched) uranium-clad in magnesium alloy or aluminum alloy, thus the word: magnox. Coolant is usually carbon dioxide gas. The magnox reactor's heat is carried along water-filled tubes by carbon dioxide. The first commercial nuclear power station opened in 1956 at Calder Hall in Cumbria, England was of this type.

Medium High Temperature Gas-Cooled Reactor [MHTGCR], graphite-moderated, helium-cooled, made of three-vessel reactor. The reactor core is composed of hexagonal blocks of graphite fuel elements in an annular array. Fuel-coated particles of Uranium Oxycarbide and Thorium Oxide, also called prismatic fuel, are used. Maximum fuel temperature can reach up to 600°C.

High Temperature Gas-Cooled Reactor [HTGCR], this is modular in design. It generated a thermal power of about 1,400 MW from four (4) nuclear modules; Electric output is about 538 MWe from two turbine-generators. Net efficiency is about 38.4%. Steam generated is about 2515 lb/in₂ at a temperature of about 1005ºF. Coolant temperature is about 1268ºF. Core power density is about 5.9/W/cm². Fuel Burn-up is about 92.2 MW·day/ton.

Advanced Gas-Cooled Reactor [AGR] uses heat exchangers located within the pressure vessel itself. The carbon dioxide coolant is pressurized and heated up to 600°C (1112°F) or more as it is pumped through the core, which is made up of fuel rods filled with enriched uranium dioxide.

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Subsections: Gas-Cooled | Light-Water | Heavy-Water | Fast Breeder | PRISM

Light Water Reactor [LWR]

Moderator: water, H₂O
Coolant: Light water, D₂O
Fuel: Uranium dioxide fabricated into pellets, right circular cylinders 19mmH x 8mm diameter enriched by 3 to 7% ²³⁵U. The pellet is made up of Uranium dioxide core, clad with two layers of pyrolitic carbon sandwiched a single layer of silicon carbide. The pellets are inserted into fuel tubes (4.5 meters overall length, with 3.8m active length) of stainless steel or Zircalloy.

Pressurized Water Reactor [PWR]. Essentially a closed lop in which the combined coolant and moderator - ordinary ("light") water - pressurized at about 150 atmosphere, and pumped through the reactor core. The core made up of fuel rods containing pellets or enriched uranium dioxide, heats the coolant to around 235°C (617°F), which then passes through a heat exchanger, where it transfers heat to a separate reservoir of water. This water is vaporized to steam, which is piped off to drive the turbine. The primary system is maintained at sub-cooled condition by operating at a pressure (12 to 16 MPa) greater than saturation. Coolant is circulated by a large pump.

Boiling Water Reactor [BWR]. The reactor vessel also contains the steam-separation apparatus, since the coolant is connected to steam in the core. Steam is piped from the reactor vessel to the turbine and condensate is returned from the hot well through feedwater systems to the reactor vessel. Usually access to the control rods are at the bottom.

There are subclasses of BWR, namely:
a. Passively Stable PWR (PS-PWR) which relies on gravity, natural circulation, and cooling by evaporation and convection rather than using pumps. An example of this is the AP-600 by Westinghouse commissioned by US DOE and the Electric Power Research Institute [EPRI]. It still uses uranium dioxide as fuel and cooling is controlled by an emergency core cooling system (ECCS) - a combination of cooling water sources: gravity drain of water and water ejected from two accumulator tanks under nitrogen pressure.
b. Passively Stable BWR (PS-BWR) which operates at full power under normal circulation, which eliminates or reduces the number of recirculation pumps.
c. Process Inherent Ultimate Safety Reactor (PIUS), or the passive LWR, a concept originating from studies of reactor systems suitable for central heating applications. PIUS is a 640-MWe PWR plant, its core is enclosed in a large prestressed concrete vessel. The fluidic valve is located at the bottom of the core. it introduces, through intrinsic thermal-hydraulic properties, emergency core cooling from the pool of water surrounding the reactor.

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Subsections: Gas-Cooled | Light-Water | Heavy-Water | Fast Breeder | PRISM

Heavy Water Reactor [HWR]

Moderator: Heavy Water, D₂0
Coolant: Heavy Water, D₂0
Fuel: Slightly enriched uranium oxide, ²³⁵U enriched by 3%.
This technology is mainly developed in Canada. Its main advantage is that it is the least wasteful of neutrons and employs a lesser number of circulating pumps. Present development includes the CANDU, or the Canadian Deuterium-Uranium Reactor. A key feature is that it is the world's only design that has two separate and independent safety shutdown systems.¹ Unlike PWR reactors, a CANDU reactor can be fuelled without shutting down the reactor. It has also a very flexible fuel cycle. It can use low-fissile fuel, such as inexpensive natural uranium or thorium. A CANDU can also use slightly enriched or recovered uranium from conventionally processed PWR fuel. This means recycled used fuel from a PWR reactor can be used in a CANDU, thus reducing the amount of used nuclear fuel waste needed to generate electricity. The CANDU, designed by the Atomic Energy of Canada Limited, reactor uses a highly stabilized fuel bundle that is easy to fabricate and manufacture. A CANDU fuel bundle, either fresh or used, has no criticality potential outside of the reactor. The design has several small diameter pressure tubes, which can be replacde, thus extending reactor life, versus the single large pressure vessel of the PWR design.

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Subsections: Gas-Cooled | Light-Water | Heavy-Water | Fast Breeder | PRISM

Fast Breeder Reactor [FBR]

Moderator: None
Coolant: Liquid metal, Sodium [Na] or Sodium Potassium [NaK]
Fuel: Plutonium Dioxide PuO₂or Uranium Dioxide, UO₂
It is called fast breeder because of fast neutron activity because it runs at a very high temperature. The heat from these is removed by a liquid metal, and two heat exchangers are needed. The hot sodium around the core first heats a different lot of sodium, which in turn generates steam.

There are reactors belonging to this class which are under research and development as of the present, namely:
a. Power Reactor Inherently Safe Module (PRISM) A pool reactor with annular flow. The sodium is circulated through the core by four cartridge-type electromagnetic pumps. In US design, fuel is a mixture of Uranium-Plutonium-Zirconium alloy with pins of Plutonium enriched at 25%. In French design, the fuel is of Uranium Dioxide pellets. Other experimental reactors use mixed oxides of Uranium or Plutonium, or mixed carbide or nitride metals. Thermal capacity is about 4239 MW using nine modules. ELectrical capacity can reach up to 1395 MWe, using three turbine-generator sets at 33% efficiency. A commercial PRISM plant is envisioned to consist of a series of three such 465-MWe power packs, each of which would be functionally independent of the other two.
b. Integral Fast Reactor (IFR) uses metallurgical processing method for the separation of Uranium, Plutonium and transuranic elements from fission products. Transuranic elements are elements above Uranium containing isotopes in the periodic table, such as ²³⁹Pu, ²⁴¹Americium, ²⁴³Americium, ²⁴⁴Cirium, ²³⁷Neptunium, and ²⁴⁵Curium.
c. European Fast Reactor (EFR) is the incorporation of the best features of the various national designs such as France's Supephenix 2, U.K.'s CDFR, and Germany's SNR 2. It retains the basic characteristics of the previous national designs, namely sodium as the coolant, a pool configuration; and oxide fuel. The EFR is specified to be the lead plant of a commercial series for 2010 onward with the technology remaining valid in 2010 without substantial extrapolation of the technology. In 1991, during the International Conference on Fast Reactors and Related Fuel Cycles, FR'91, The envisaged size of 1500 MWe is intended to take advantage of the economy of scale. Reactor heat output is about 3600 MWt. Net electrical output at about 1440 MWe, with a 40% efficiency. The core will have two enrichment zones with a fuel management scheme being based on a six-year residence time. The Primary system uses six intermediate heat exchangers and three primary pumps. Heat is transported from the primary system to the steam circuits by a secondary system consisting of six intermediate heat exchangers with six steam generators. The decay heat removal concept minimizes the dependence on safety-graded emergency power supplies. Six dip coolers in the primary vessel are connected to six sodium air coolers arranged in high stacks. The main pipes connecting the components have low flow resistance to enhance natural circulation. The goal is to enhance passive capabilities by natural circulation of the sodium. The entire nuclear island will rest on a single foundation, and all buildings will be seismically isolated by rubber isolators tuned to a frequency of about 1 Hz.

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1. Author. TitleBook Title Publisher, Place, Year, page.

1. Allen Kirkpatrick, Nuclear Power: Looking for a leading role., Asian Energy Infrastructure, December 2000, Volume 2, Issue 4, PennWell Corporation, USA p. 30

Synthetic Gas

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