There has been a revival of interest in small and simpler nuclear units for generating electricity and for process heat. This interest is being driven both by a desire to reduce capital costs and to provide power away from the main grid systems.
Since nuclear power became established in the 1950s, the size of reactors has grown from 60 MWe to over 1300 MWe, with corresponding economies of scale in operation. At the same time, there have been hundreds of smaller reactors built for naval use (up to 190 MWt) and as neutron sources, resulting in considerable experience in the design and construction of deliberately small reactors.
Today, partly due to the high capital cost of large power reactors generating electricity via a steam cycle, and partly due to consideration of public perception, 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. The numbers produced provides economies of scale. There are also moves to develop small units for remote sites. The International Atomic Energy Agency (IAEA) defines small as under 300 MWe.
Figure 1. The PBMR power conversion unit is based on the thermodynamic Brayton (gas turbine) cycle
The most prominent modular project is the South African-led consortium developing the Pebble Bed Modular Reactor (PBMR) of 110 MWe. There is also a design for a 285 MWe version being developed. Both drive gas turbines directly, using helium as a coolant and operating at very high temperatures. They build on the experience of several innovative reactors in the 1960s and 1970s.
Generally, modern small reactors for power generation are expected to have simpler designs, economy of mass production, and reduced siting costs. Many are also designed for a high level of passive or inherent safety in the event of malfunction.
Some are designed 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 50 MWe unit is estimated to be 5.4-10.7 à‚¢/kWh, compared with charges in Alaska and Hawaii of 5.9-36 à‚¢/kWh.
US Congress is funding research on both small modular nuclear power plants assembled on site from factory-produced modules, and advanced gas-cooled designs, which are modular in the sense that ten or more units are progressively built to comprise a major power station. A report from the US Department of Energy in 2001 considered nine designs that could possibly be deployed by 2010.
In Siberia, there are four small units already operating at the Bilibino cogeneration plant. These four 62 MWt units are an unusual graphite-moderated boiling water design with water/steam channels through the moderator. They produce steam for district heating and 11 MWe each. They have performed well since 1976, and are much cheaper than fossil fuel alternatives in the Arctic region.
Light water reactors
US experience has been of very small power plants, such as the 11 MWt, 1.5 MWe PM-3A reactor that operated at McMurdo Sound in Antarctica from 1962-1972, generating a total of 78 GWh. There was also an Army programme for small reactor development, and some successful small reactors from the main national programme that started in the 1950s. One was the 67 MWe Big Rock Point BWR (Boiling Water Reactor) that operated for 35 years to 1997.
Figure 2. The CAREM advanced nuclear power plant is being developed by CNEA and INVAP
The Russian KLT-40 is a reactor that has been well-proven in icebreakers, and is now proposed for wider use in desalination and on barges for remote area power supply where it produces 35 MWe as well as heat. While these are designed to run for three years between refuelling, it is envisaged that they will be operated in pairs to allow for outages, perhaps with on-board refuelling capability and spent fuel storage.
Although the reactor core is normally cooled by forced circulation, the design relies on convection for emergency cooling. Fuel is a U-Al alloy with burnable poison, clad with zircaloy, and may be highly enriched. Up to 35 MWt can be utilized for desalination in addition to the electrical output.
The CAREM advanced small nuclear power plant being developed by CNEA and INVAP in Argentina is a modular
100 MWt/25MWe pressurised water reactor (PWR) with integral steam generators designed to be used for electricity generation or as a research reactor or for water desalination. CAREM has its entire primary coolant system within the reactor pressure vessel, self-pressurized and relying entirely on convection. Fuel is standard 3.4 per cent enriched PWR fuel, with burnable poison, and is refuelled annually. It is a mature design that could be deployed within a decade.
Figure 3. The GT-MHR is being developed by General Atomics in partnership with Russia’s Minatom, supported by Framatome ANP and Fuji of Japan. Initially, it will be used to burn pure ex-weapons plutonium at Tomsk in Russia
On a larger scale, South Korea’s System-Integrated Modular Advanced Reactor (SMART) is a 330 MWt PWR, with integral steam generators and advanced safety features. It is designed for generating electricity (up to 100 MWe) and/or thermal applications such as desalination. The design life is 60 years, with a three-year refuelling cycle. A one-fifth scale plant (65 MWt) is being constructed, with first operation due in 2007.
The Japan Atomic Energy Research Institute (JAERI) is developing the MRX, a small (50-300 MWt) integral PWR for marine propulsion or local energy supply (30 MWe). The entire plant would be factory built. It has conventional 4.3 per cent enriched PWR uranium oxide fuel with a 3.5 year refuelling interval. It also has a water-filled containment to enhance safety. It could be deployed within a decade.
Westinghouse is developing the International Reactor Innovative and Secure (IRIS) as a Generation IV project. IRIS-50 is a modular 50 MWe PWR with integral primary coolant system and circulation by convection. Fuel is similar to present Light Water Reactors (LWR). Enrichment is five per cent with burnable poison and a refuelling interval of five 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 US at both 200 MWe and 50 MWe levels. The design is based on GE’s SBWR. It uses convection in the coolant and has a five per cent enriched BWR fuel with a ten-year refuelling interval. It may be ready for deployment this decade.
High temperature designs
These reactors use helium as a coolant that at up to 950à‚°C drives a gas turbine to generate electricity and a compressor to return the gas to the reactor core. Fuel is in the form of particles less than 1 mm in diameter. Each has a kernel of uranium oxycarbide, with the uranium enriched up to eight per cent U-235. This is surrounded by layers of carbon and silicon carbide, giving a containment for fission products that is stable to 2000à‚°C. These particles may be arranged in hexagonal prisms of graphite, or in pebbles of graphite encased in silicon carbide the size of billiard balls.
A consortium led by the South African utility Eskom is developing South Africa’s PBMR with a direct-cycle gas turbine generator. Modules will be 110 MWe, and thermal efficiency will be 45 per cent. Up to 450 000 fuel pebbles cycle through the graphite-lined reactor continuously, about ten times each, until they are expended, giving an average enrichment in the fuel load of five to six per cent and burn-up of 80 000 MWday/tU. Eventual target burn-ups are 200 000 MWd/tU. Control rods are in the side reflectors. Performance is said to include great flexibility in loads, with rapid change in power settings. Each unit will finally discharge about 19 t/year of spent pebbles to ventilated on-site storage bins.
Construction cost for 10-14 units is expected to be $1000/kW, and generating costs 1.6 à‚¢/kWh. A prototype is due to enter operation in 2006.
A large US design, the Gas Turbine ” Modular Helium Reactor (GT-MHR) ” will be built as 285 MWe modules, each directly driving a gas turbine at 48 per cent 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. It is being developed by General Atomics in partnership with Russia’s Minatom, supported by Framatome ANP and Fuji of Japan. Initially, it will be used to burn pure ex-weapons plutonium at Tomsk in Russia. Plant costs are expected to be less than $1000/ kW.
Figure 4. High temperatures mean high thermal efficiency in the GT-MHR project
A smaller version of this, the Remote Site Modular Helium Reactor (RS-MHR) of 10-25 MWe has been proposed by General Atomics. The fuel would be 20 per cent enriched, and refuelling interval would be six to eight years.
Liquid metal-cooled reactors
The encapsulated nuclear heat source (ENHS) is a liquid metal-cooled 50 MWe reactor being developed by the University of California. The core is in a metal-filled module sitting in a large pool of secondary molten metal coolant that also accommodates the separate and unconnected steam generators. The fuel is a uranium-zirconium alloy with 13 per cent uranium enrichment, or U-Pu-Zr with 11 per cent Pu, with a 15-year life. After this, the module is removed, stored on site until the primary lead or Pb-Bi coolant solidifies, and it would then be shipped as a self-contained and shielded item. A new-fuelled module would be supplied complete with primary coolant. The ENHS is designed for developing countries, but it is not yet close to commercialization.
A related project is the secure transportable autonomous reactor for hydrogen production ” STAR H2. It is a 400 MWt lead-cooled fast neutron modular reactor with passive safety features. Its size means that it can be shipped by rail and cooled by natural circulation. It uses U-Transuranic nitride fuel in a cassette that is replaced every 15 years. The reactor heat at 780à‚°C is conveyed by a helium circuit to drive a separate thermochemical hydrogen production plant, while lower grade heat is harnessed for desalination using the multi-stage flash process. Any commercial electricity generation would be using the hydrogen in fuel cells.
For both of 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 burn-up of uranium and transuranics is envisaged in STAR-H2, with only fission products being waste. Russia has experimented with several lead-cooled reactor designs, and has used lead-bismuth cooling for 40 years in its submarine reactors. Pb-208 (54 per cent of naturally occurring lead) is transparent to neutrons. A significant Russian design is the 300 MWe BREST fast neutron reactor, with lead as the primary coolant at 540à‚°C, and supercritical steam generators. It is inherently safe and uses a U+Pu nitride fuel. No weapons-grade plutonium can be produced, as there is no uranium blanket, and spent fuel can be recycled indefinitely with on-site facilities. A pilot unit is being built at Beloyarsk, and 1200 MWe units are planned.
Japan’s Central Research Institute of Electric Power Industry (CRIEPI) is developing the 50 MWe 4S or Rapid-A system. It uses sodium as coolant and has passive safety features. The whole unit would be factory built. Fuel is uranium-zirconium alloy enriched to 15 per cent, with refuelling every ten years. Steady power output is achieved by progressively withdrawing a graphite reflector around the slender core. It is unlikely to be in use before 2010.
A small-scale design funded by the Japan Atomic Energy Research Institute (JAERI) is the 200 kWe Rapid-L, which uses lithium-6 as control medium. It would have 2700 uranium nitride fuel pins with 2600à‚°C melting point integrated into a disposable cartridge. The reactivity control system is passive, using lithium expansion modules that give burn-up compensation, partial load operation as well as negative reactivity feedback. As the reactor temperature rises, the lithium expands into the core, displacing an inert gas. Other lithium modules, also integrated into the fuel cartridge, shut down and start up the reactor. Cooling is by molten sodium. Refuelling would be every five years in an inert gas environment. Operation would require no skill, due to the inherent safety design features. The whole plant would be 6.5 m high and 2 m diameter.
Molten salt reactors
During the 1960s, the US 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 US.
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 700à‚°C. 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 so.
The attractive features of the MSR fuel cycle include: the high-level waste comprising fission products only, hence shorter-lived radioactive waste, small inventory of weapons-fissile material, low fuel use and safety due to passive cooling up to any size.
A US development of the MSR uses a graphite matrix fuel similar to that in HTGRs and with a similar fuel cycle to them. The salt, with better heat transfer properties than helium, is used solely as coolant, and achieves temperatures of 1000à‚°C while at low pressure. This could be used in thermochemical hydrogen manufacture.