As nuclear power generation has developed since the 1950s, the size of reactors has grown from 60MWe to over 1300MWe, with corresponding economies of scale in operation. At the same time, there have been many hundreds of smaller reactors built for naval use and as neutron sources.

Today, partly due to the high capital cost of large power reactors, there is a move to develop smaller units. These may be built independently or as modules in a larger complex, with capacity added incrementally as required. Economies of scale are provided by the numbers produced. There are also moves to develop small units for remote sites.

Generally, modern small reactors for power generation are expected to have simpler design, the economics of mass production and reduced siting costs. Many are also designed for a high level of passive safety in the event of malfunction.

Some are conceived for areas away from transmission grids and with small loads. Others are designed to operate in clusters in competition with large units. The cost of electricity from a 50MWe unit is estimated by the US Department of Energy (DoE) as 5.4-10.7¢/kWh (compared with charges in Alaska and Hawaii of 5.9-36.0¢/kWh).

The US Congress is now funding research on both small modular nuclear power plants and advanced gas-cooled designs. A DoE report in 2001 considered nine designs which could be deployed by 2010.

In a remote corner of Siberia, four small units have operated at at the Bilibino cogeneration plant since 1976. These four 62MWt units are an unusual graphite-moderated boiling water design with water/steam channels through the moderator. They produce steam for district heating and power – 11MWe each.

Light water reactors

The Russian KLT-40 is a reactor well proven in icebreakers, and it is currently being proposed for wider use in desalination and on barges for power supply to remote areas. The reactors are designed to run for 3 years between refuelling, and it is suggested that they be operated in pairs to allow for outages, perhaps with on-board refuelling capability and spent fuel storage.

Although the KLT-40 reactor core is normally cooled by forced circulation, the design relies on convection for emergency cooling. The fuel is U-Al alloy with burnable poison, clad with zircalloy, and may be highly enriched. Up to 35MWt can be utilised for desalination in addition to the 40MW of electrical output.

The Carem advanced small nuclear power plant being developed by CNEA and INVAP in Argentina is a modular 100MWt/25MWe PWR with integral steam generators designed to be used for electricity generation, as a research reactor or for desalination. Carem has its entire primary coolant system within the reactor pressure vessel, is self-pressurised and relies entirely on convection. Fuel is standard 3.4%-enriched PWR fuel, with burnable poison, and is replaced annually. It is a mature design which could be deployed within a decade.

On a larger scale, South Korea’s Smart (System-integrated Modular Advanced Reactor) is a 330MWt PWR with integral steam generators and advanced safety features. It is designed for generating electricity (up to 100MWe) and/or thermal applications such as seawater desalination. Smart’s design life is 60 years, with a 3-year refuelling cycle. A one-fifth scale plant (65MWt) is being constructed for operation by 2007.

The Japan Atomic Energy Research Institute (JAERI) is developing the MRX, a small (50-300MWt) integral PWR for marine propulsion or local energy supply (30MWe). The entire plant would be factory-built. It has conventional 4.3%-enriched PWR uranium oxide fuel with a 3.5 year refuelling interval. It has a water-filled containment to enhance safety. It could be deployed within a decade.

The International Reactor Innovative and Secure (IRIS) is being developed by Westinghouse as a Generation IV project. IRIS-50 is a modular PWR of 50MWe or more. It will have an integral primary coolant system with circulation by convection. The fuel is similar to existing LWRs, enriched to 5% with burnable poison and a fuelling interval of 5 years, or longer with higher enrichment. IRIS-50 could be deployed this decade.

The Modular Simplified Boiling Water Reactor (MSBWR) is being developed by General Electric and Purdue University in the USA at both 200MWe and 50MWe sizes, based on GE’s SBWR. It is convection-cooled using 5%-enriched BWR fuel with a 10-year refuelling interval. It may be ready for deployment this decade.

The TRIGA Power System is a PWR concept based on General Atomics’ research reactor design. It is conceived as a 64MWt, 16.4MWe pool-type system operating at a relatively low temperature. The secondary coolant is organic perfluorocarbon. The fuel is uranium-zirconium hybrid enriched to 20% with a little burnable poison. It has an 18-month refuelling cycle with spent fuel stored inside the reactor vessel.

High temperature

gas-cooled reactors

High temperature gas cooled reactors use helium coolant at up to 950ºC to drive a gas turbine to produce electricity, and a compressor to return the gas to the reactor core.

The fuel is in the form of particles less than a millimetre in diameter. Each has a kernel of uranium oxycarbide, with the uranium enriched up to 8%. This is surrounded by layers of carbon and silicon carbide, giving a containment for fission products which is stable to 2000ºC. These particles may be arranged in hexagonal blocks of graphite, or in billiard ball-sized pebbles of graphite encased in silicon carbide. Several reactor designs are in development and there is considerable basic research also under way.

The Pebble Bed Modular Reactor (PBMR), with a direct-cycle gas turbine generator, is being developed by a consortium being led by Eskom. Modules will be of 120MWe and thermal efficiency of about 45%. Up to 450,000 fuel pebbles recycle through the graphite-lined reactor continuously (about ten times each) until they are expended, giving an average enrichment in the fuel load of 5-6% and burn-up of 80 GWday/tU (target burn-ups are 200GWd/tU). Control rods are in the side reflectors. Each unit will discharge about 19t/yr of spent fuel to ventilated on-site storage.

Construction costs for clusters of 10-14 units are expected to be $1000/kWe, with operating costs of 1.6¢/kWh. A prototype was due to be built in 2002, but the design feasibility study has been delayed and is still underway.

A larger US design, the Gas Turbine Modular Helium Reactor (GT-MHR), will be built as modules of 285MWe, each directly driving a gas turbine at 48% thermal efficiency. The cylindrical core consists of 102 hexagonal fuel element columns of graphite blocks with channels for helium and control rods. Graphite reflector blocks are both inside and around the core. Half the core is replaced every 18 months.

The GT-MHR is being developed by General Atomics in partnership with Russia’s Minatom, supported by Framatome ANP and Fuji. Initially GT-MHR will be used to burn pure ex-weapons plutonium at Tomsk. Plant costs are expected to be less than $1000/kWe.

General Atomics has also proposed a Remote Site Modular Helium Reactor (RS-MHR) of 10-25MWe. The fuel would be 20% enriched and refuelling intervals would be 6-8 years.

Liquid metal cooled reactors

The Encapsulated Nuclear Heat Source (ENHS) is a 50MWe liquid metal-cooled reactor being developed at the University of California. The core is in a metal-filled module sitting in a large pool of secondary molten metal coolant, which also contains the steam generators. Fuel is a uranium-zirconium alloy with 13% enrichment (or U-Pu-Zr with 11% Pu). At the end of its 15-year life, the module is removed, replaced with a module complete with coolant, and stored on site until the primary lead coolant solidifies. It is then shipped as a self-contained and shielded item. Designed for developing countries, ENHS is not yet close to commercialisation.

A related project is the secure transportable autonomous reactor for hydrogen production (STAR-H2). It is a lead-cooled fast neutron modular reactor with passive safety features. Its 400MWt size means that it can be shipped by rail and cooled by natural circulation. It uses U-transuranic nitride fuel in an assembly which is replaced every 15 years.

The reactor heat at 780ºC is conveyed by a helium circuit to a separate thermochemical hydrogen production plant, while lower grade heat is harnessed for desalination (multi-stage flash process).

In both these concepts regional fuel cycle support centres would handle fuel supply and reprocessing, and fresh fuel would be spiked with fission products to deter misuse. Complete burnup of uranium and transuranics is envisaged in STAR-H2, so the waste comprises only fission products. Russia has experimented with several lead-cooled reactor designs, and has used lead-bismuth cooling for 40 years in its submarine reactors.

A significant Russian design is the BREST 300MWe lead-cooled fast neutron reactor. It uses a U+Pu nitride fuel and no weapons-grade Pu can be produced, as there is no uranium blanket. A pilot unit is being built at Beloyarsk and 1200MWe units are planned.

The 50MWe 4S or Rapid-A sodium-cooled system is being developed by Japan’s Central Research Institute of Electric Power Industry (CRIEPI). The whole unit would be factory-built. Fuel is 15%-enriched uranium-zirconium alloy, refuelled every ten years. Steady power output is achieved by progressively withdrawing a graphite reflector around the slender core. It is unlikely to be used by 2010.

A small-scale design from the same Japanese source, but funded by the Japan Atomic Energy Research Institute (JAERI) is the 200kWe Rapid-L, which uses Li-6 as control medium.

Japan’s LSPR is a lead-bismuth cooled reactor of 150MWt/53MWe. Fuelled units would be supplied from a factory and operate for 30 years, and then be returned. The design is intended for use in developing countries.

In the USA, GE was involved in designing a modular 150MWe liquid metal-cooled inherently safe reactor known as PRISM.

Molten salt reactors

During the 1960s, the USA developed the molten salt breeder reactor as the primary back-up option for the fast breeder reactor (cooled by liquid metal), and a small prototype was operated. There is now renewed interest in the concept in Japan, Russia, France and the USA.

In a molten salt reactor (MSR), the fuel is a molten mixture of lithium and beryllium fluoride salts with dissolved thorium and U-233 fluorides. The core consists of unclad graphite arranged to allow the flow of salt at some 700ºC. Heat is transferred to a secondary salt circuit and thence to steam. The fission products dissolve in the salt and are removed continuously in an on-line reprocessing loop and replaced with Th-232 or U-238. Actinides remain in the reactor until they fission or are converted to higher actinides which do.

The attractive features of the MSR fuel cycle include the high-level waste comprising only fission products. This leads to shorter-lived radioactivity; a small inventory of weapons-fissile material (Pu-242 being the dominant Pu isotope); low fuel use (the French self-breeding variant claims 50kg of thorium and 50kg U-238 per billion kWh); and safety due to passive cooling up to any size.

A US development of the MSR uses graphite matrix fuel similar to that in HTGRs and with a similar fuel cycle. The salt, with better heat transfer properties than helium, is used solely as coolant, and achieves temperatures of 1000ºC while at low pressure. This could be used in thermochemical hydrogen manufacture.
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