The nuclear power industry has been developing and improving reactor technology for almost five decades and is preparing for the next generations of nuclear power reactors to fill orders expected in the next five to twenty years.

Several generations of reactors are commonly distinguished. Generation I reactors were developed in 1950-60s and outside the UK none are still running today. Generation II reactors are typified by the present US fleet and most in operation elsewhere. Generation III are the Advanced Reactors discussed in this paper; the first are in operation in Japan and others are under construction or ready to be ordered. Generation IV reactor designs are still on the drawing board and will not be operational before 2020 at the earliest.

About 85% of the world's nuclear electricity is generated by reactors derived from designs originally developed for naval use. These and other second-generation nuclear power units have been found to be safe and reliable, but they are being superseded by better designs.

Reactor suppliers in North America, Japan, Europe, Russia and South Africa have a dozen new nuclear reactor designs at advanced stages of planning, while others are at a research and development stage. Fourth-generation reactors are at concept stage.

Third-generation reactors have:

• a standardized design for each type to expedite licensing, reduce capital cost and reduce construction time;
• a simpler and more rugged design, making them easier to operate and less vulnerable to operational upsets;
• higher availability and longer operating life—typically 60 years;
• reduced possibility of core melt accidents;
• minimal effect on the environment;
• higher burn-up to reduce fuel use and waste; and
• burnable absorbers ("poisons") to extend fuel life.

The greatest departure from second-generation designs is that many third-generation reactors incorporate passive or inherent safety features that require no active controls or operational intervention to avoid accidents in the event of malfunction, and may rely on gravity, natural convection or resistance to high temperatures. Traditional nuclear reactor safety systems are 'active' in the sense that they involve electrical or mechanical operation on command. Some engineered systems operate passively, e.g., pressure relief valves. Both require parallel redundant systems. Inherent or full passive safety depends only on physical phenomena such as convection, gravity or resistance to high temperatures, not on functioning of engineered components. Many are larger than their predecessors.

Joint Initiatives

Two major international initiatives have been launched to define future nuclear reactor and nuclear fuel cycle technology, mostly looking further ahead than the main subjects of this paper:

• Generation IV International Forum (GIF) is a US-led grouping set up in 2001 that has identified six reactor concepts for further investigation with a view to commercial deployment by 2030.

• The International Atomic Energy Agency (IAEA)'s International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) is focused more on developing country needs, and initially involved Russia rather than the USA, though the USA has now joined it. It is now funded through the IAEA budget.

Light Water Reactors (LWR)

In the USA, the federal Department of Energy (DOE) and the commercial nuclear industry in the 1990s developed four advanced reactor types. Two of them fall into the category of large "evolutionary" designs which build directly on the experience of operating light water reactors in the USA, Japan and Western Europe. These reactors are in the 1300 megawatt range.

One is an advanced boiling water reactor (ABWR), four examples of which are in commercial operation in Japan, with another under construction there and two in Taiwan. Four more are planned in Japan and another in the USA. The other type, System 80+, is an advanced pressurized water reactor (PWR), which was ready for commercialization but is not now being promoted for sale. Eight System 80 reactors in South Korea incorporate many design features of the System 80+, which is the basis of the Korean Next Generation Reactor program, specifically the APR-1400, which is expected to be in operation soon after 2010 and marketed worldwide. The US Nuclear Regulatory Commission (NRC) gave final design certification for both in May 1997, noting that they exceeded NRC "safety goals by several orders of magnitude". The ABWR has also been certified as meeting European requirements for advanced reactors (see below).

Another, more innovative, US advanced reactor is smaller - 600 MWe - and has passive safety features (its projected core damage frequency is nearly 1000 times less than today's NRC requirements). The Westinghouse AP-600 gained NRC final design certification in 1999 (AP = Advanced Passive).

These NRC approvals were the first such generic certifications to be issued and are valid for 15 years. As a result of an exhaustive public process, safety issues within the scope of the certified designs have been fully resolved and hence will not be open to legal challenge during licensing for particular plants. Utilities will be able to obtain a single NRC licence to both construct and operate a nuclear reactor before construction begins.

Separate from the NRC process and beyond its immediate requirements, the US nuclear industry selected one standardized design in each category—the large ABWR and the medium-sized AP-600, for detailed first-of-a-kind engineering (FOAKE) work. The US\$200 million program, half funded by DOE, is now complete. It means that prospective buyers now have firm information on construction costs and schedules.

The Westinghouse AP-1000, scaled-up from the AP-600, has now received final design approval from the NRC and is expected to gain full design certification in 2005. It represents the culmination of a 1300 man-year and \$440 million design and testing program. Overnight capital costs are projected at \$1200 per kilowatt and modular design will reduce construction time to 36 months. The 1100 MWe AP-1000 generating costs are expected to be below US\$3.5 cents/kWh and it has a 60 year operating life. It is under active consideration for building in China, Europe and the USA, and is capable of running on a full MOX core if required.

General Electric has developed the ESBWR of 1390 MWe with passive safety systems from its ABWR design. This then grew to 1550 MWe and has been submitted for NRC design certification in the USA. Design approval is expected by 2007. It is favored for early US construction.

Another US-origin but international project, a few years behind the AP-1000, is the International Reactor Innovative & Secure (IRIS) project. Westinghouse is leading a wide consortium developing it as an advanced 3rd Generation project. IRIS is a modular 335 MWe pressurized water reactor with integral steam generators and primary coolant system all within the pressure vessel. Fuel is initially similar to present LWRs with 5% enrichment and burn-up of 60 GWd/t with fuelling interval of up to 4 years, but is designed ultimately for 10% enrichment and 80,000 MWd/t burn-up with an 8 year cycle, or equivalent MOX core. IRIS could be deployed in the next decade, and US design certification is envisaged by 2010. Multiple modules are expected to cost US\$1000-1200 per kilowatt.

In Japan, the first two ABWRs, Kashiwazaki Kariwa-6 & 7, have been operating since 1996 and are expected to have a 60 year life. These cost about US\$2000/kW to build, and produce power at about US\$7 cents/kWh. A third unit started up in 2004. Future ABWR units are expected to cost US\$1700/kW. Several 1350 MWe units are under construction in Japan and Taiwan.

To complement this ABWR, Hitachi has completed systems design for three more of the same type—600, 900 and 1700 MWe versions of the 1350 MWe design. The smaller versions will have standardized features which reduce costs. Construction of the ABWR-600 is expected to take 34 months—significantly less than the 1350 MWe units.

A large (1500 MWe) advanced pressurized water reactor (APWR) is being developed by four utilities together with Westinghouse and Mitsubishi. The first two are planned for Tsuruga. The design is simpler, combines active and passive cooling systems to greater effect, and has over 55 GWd/t fuel burn-up. Design work continues and will be the basis for the next generation of Japanese PWRs. In addition, Mitsubishi is participating in development of Westinghouse's AP-1000 reactor.

In South Korea, the APR-1400 Advanced PWR design has evolved from the US System 80+ with enhanced safety and seismic robustness and was earlier known as the Korean Next-Generation Reactor. Design certification by the Korean Institute of Nuclear Safety was awarded in May 2003. The first of these 1450 MWe reactors will be Shin-Kori-3 & 4, expected to be operating by 2012. The fuel has burnable poison and will have up to 60 GWd/t burn-up. Projected cost is US\$1400 per kilowatt, falling to \$1200/kW in later units with 48-month construction time. Plant life is expected to be 60 years.

In Europe, four designs are being developed to meet the European Utility Requirements (EUR) of French and German utilities, which have stringent safety criteria.

Framatome ANP has developed a large (1600 and up to 1750 MWe) European pressurized water reactor (EPR), which was confirmed in mid-1995 as the new standard design for France and received French design approval in 2004. It is derived from the French N4 and German Konvoi types and is expected to provide power about 10% cheaper than the N4. It will operate flexibly to follow loads, have fuel burn-up of 65 GWd/t and the highest thermal efficiency of any light water reactor, at 36%. Availability is expected to be 92% over a 60-year service life. The first unit is set to be built at Olkiluoto in Finland, the second at Flamanville in France. A US version of the EPR is also undergoing review in USA with intention of a design certification application in 2007.

Together with German utilities and safety authorities, Framatome ANP is also developing another evolutionary design, the SWR 1000, a 1000-1290 MWe BWR. The design was completed in 1999 and development continues, with US design certification being sought. Along with many passive safety features, the nuclear reactor design is simpler overall and uses high-burnup fuels, giving it refuelling intervals of up to 24 months. This design is ready for commercial deployment.

General Electric has developed the European Simplified Boiling Water Reactor of 1390 MWe with passive safety systems, from its ABWR design. This is now known as the Economic & Simplified BWR (ESBWR); a 1500 MWe version is at pre-application stage for NRC design certification in the USA.

Westinghouse has been developing its evolutionary BWR 90+ (1500 MWe) design, working with Scandinavian utilities to meet EUR requirements.

In Russia, several advanced reactor designs have been developed—advanced PWR with passive safety features.

The Gidropress 1000 MWe V-392 (advanced VVER-1000) units are planned for Novovoronezh and are being built in India. A transitional VVER-91 (1000 MWe) was developed with western control systems—two are being built in China at Jiangsu Tianwan, and it was bid for Finland.

The VVER-1500 V-448 model is being developed by Gidropress, and two units each are planned as replacement plants for Leningrad and Kursk. It will have 50-60 MWd/t burn-up and enhanced safety. Design is expected to be complete in 2007 and the first units commissioned in 2012-13.

Two 640 MWe PWR reactors have been developed, one by Gidropress with Siemens control systems is the VVER-640 or V-407. The other is VPBER-600 designed by OKBM with integral steam generators. Both have enhanced safety.

Heavy Water Reactors

In Canada, Atomic Energy of Canada Limited (AECL) has had two designs under development that are based on its reliable CANDU-6 (CANada Deuterium Uranium) nuclear reactors, the most recent of which are operating in China.

The CANDU-9 (925-1300 MWe) was also developed from the CANDU-6 as a single-unit plant. It has flexible fuel requirements ranging from natural uranium through slightly-enriched uranium, recovered uranium from reprocessing spent pressurized water reactor (PWR) fuel, mixed oxide (MOX or U & Pu) fuel, direct use of spent PWR fuel, to thorium. It may be able to burn military plutonium or actinides separated from reprocessed PWR/BWR waste. A two-year licensing review of the CANDU-9 design was successfully completed in early 1997.

Some innovations gained from this design, along with experience in building recent Korean and Chinese units, was then put back into the Enhanced CANDU-6—built as twin units—with a power increase to 750 MWe and flexible fuel options, plus 4.5-year construction and 60-year plant life.

The Advanced Candu Reactor (ACR), a third-generation reactor, is more of a innovative concept. While retaining the low-pressure heavy water moderator, it incorporates some features of the pressurized water reactor. Adopting light water cooling and a more compact core reduces capital cost, and because the reactor is run at higher temperature and coolant pressure, it has higher thermal efficiency.

The ACR-700 is a 750 MWe design but is physically much smaller, simpler and more efficient as well as 40% cheaper than the CANDU-6. But the ACR-1000 of 1200 MWe is now the focus of attention by AECL. It has more fuel channels—each of which can be regarded as a module of about 2.5 MWe. Projected overnight capital cost of US\$1000/kWe and operating costs of US\$3 cents/kWh have been claimed. The ACR will run on low-enriched uranium (about 1.5-2.0% uranium-235) with high burn-up, extending the fuel life by about three times and reducing high-level waste volumes accordingly. Regulatory confidence in safety is enhanced by negative void reactivity for the first time in CANDU, as well as by utilizing other passive safety features. Units will be assembled from prefabricated modules, reducing construction time to 3.5 years.

ACR is moving towards design certification in Canada, with a view to following in China, USA and UK. The first ACR-1000 unit is expected to be operating by 2014 in Ontario.

The CANDU X is a variant of the ACR, but with supercritical light water coolant (e.g., 25 MPa and 625°C) to provide 40% thermal efficiency. The size range envisaged is 350 to 1150 MWe, depending on the number of fuel channels used. Commercialization envisaged after 2020.

India is developing the Advanced Heavy Water reactor (AHWR) as the third stage in its plan to utilize thorium to fuel its overall nuclear power program. The AHWR is a 300 MWe nuclear reactor moderated by heavy water at low pressure. The calandria has 500 vertical pressure tubes and the coolant is boiling light water circulated by convection. Each fuel assembly has 30 thorium-uranium-233 oxide pins and 24 plutonium-thorium oxide pins around a central rod with burnable absorber. Burn-up of 24 GWd/t is envisaged. It is designed to be self-sustaining in relation to uranium-233 bred from thorium-232 and have a low plutonium inventory and consumption, with slightly negative void coefficient of reactivity.

High-Temperature Gas-Cooled Reactors

Nuclear fuel pebbbles used in High Temperature Gas Cooled Reactors (Image: South African Pebble Bed Modular Reactor Company)

These reactors use helium as a coolant, which, at up to 950°C drives a gas turbine for electricity-generation and a compressor to return the gas to the reactor core. Fuel is in the form of TRISO particles less than a millimeter in diameter. Each has a kernel of uranium oxycarbide, with the uranium enriched up to 17% uranium-235 (²³⁵U). This is surrounded by layers of carbon and silicon carbide, giving a containment for fission products which is stable to 1600°C or more. These particles may be arranged in blocks as hexagonal 'prisms' of graphite, or in billiard ball-sized pebbles of graphite encased in silicon carbide.

South Africa's Pebble Bed Modular Reactor (PBMR) is being developed by a consortium led by the utility Eskom, and drawing on German expertise. It aims for significant enhancements in safety, economics and proliferation resistance. Production units will be 165 MWe. They will have a direct-cycle gas turbine generator and thermal efficiency of about 42%. Up to 450,000 fuel pebbles recycle through the reactor continuously (about six times each) until they are expended, giving an average enrichment in the fuel load of 4-5% and average burn-up of 90 GWday/t U (eventual target burn-ups are 200 GWd/t). This means on-line refuelling as expended pebbles are replaced, giving a high capacity factor. The pressure vessel is lined with graphite and there is a central column of graphite as reflector. Control rods are in the side reflectors and cold shutdown units in the center column.

Performance includes great flexibility in loads (40-100%), with rapid change in power settings. Power density in the core is about one-tenth of that in light water reactor, and if coolant circulation ceases, the fuel will survive initial high temperatures while the nuclear reactor shuts itself down—an inherent safety feature. Each unit will finally discharge about 19 tonnes/yr of spent pebbles to ventilated on-site storage bins.

Overrnight construction cost (when in clusters of eight units) is expected to be US\$1000/kW and generating cost below US\$3 cents/kWh. Investors in the PBMR project are Eskom, the South African Industrial Development Corporation, and Westinghouse. A demonstration plant is due to be built in 2006 for commercial operation in 2010.

A larger US design, the Gas Turbine-Modular Helium Reactor (GT-MHR), will be built as modules of 285 MWe 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 of the core is replaced every 18 months. Burn-up is about 100,000 MWd/t. The GT-MHR is being developed by General Atomics in partnership with Russia's Minatom, supported by Fuji (Japan). Initially, it will be used to burn pure ex-weapons plutonium at Tomsk in Russia. The preliminary design stage was completed in 2001. Overnight plant costs are expected to be around US\$1000/kW and total generating cost US\$2.9 cents/kWh.

HTRs can potentially use thorium-based fuels, such as HEU with thorium (Th), uranium-233 with Th, and plutonium with Th. Most of the experience with thorium fuels has been in HTRs.

Fast Neutron Reactors

Several countries have research and development programs for improved Fast Breeder Reactors (FBR), which are a type of Fast Neutron Reactor. These use the uranium-238 (²³⁸U) in reactor fuel as well as the fissile uranium-235 (²³⁵U) isotope used in most nuclear reactors.

About 20 liquid metal-cooled FBRs have already been operating, some since the 1950s, and some supply electricity commercially. About 290 reactor-years of operating experience have been accumulated.

Natural uranium contains about 0.7% ²³⁵U and 99.3 % ²³⁸U. In any reactor, the ²³⁸U component is turned into several isotopes of plutonium during its operation. Two of these, plutonium-239 (²³⁹Pu) and ²⁴¹Pu, then undergo fission in the same way as ²³⁵U to produce heat. In a fast neutron reactor, this process is optimized so that it can 'breed' fuel, often using a depleted uranium blanket around the core. FBRs can utilize uranium at least 60 times more efficiently than a normal reactor. They are, however, expensive to build and could only be justified economically if uranium prices were to rise to pre-1980 values, well above the current market price. For this reason, research work on the 1450 MWe European FBR has almost ceased. Closure of the 1250 MWe French Superphenix FBR after very little operation over 13 years also set back developments.

Research continues in India; at the Indira Gandhi Centre for Atomic Research, a 40 MWt fast breeder test reactor has been operating since 1985. In addition, the tiny Kamini reactor at the Centre is employed to explore the use of thorium as nuclear fuel by breeding fissile uranium-233 (²³³U). In 2004, construction of a 500 MWe prototype fast breeder reactor started at Kalpakkam. The unit is expected to be operating in 2010, fuelled with uranium-plutonium carbide (the reactor-grade Pu being from its existing PHWRs) and with a thorium blanket to breed fissile ²³³U. This will take India's ambitious thorium program to stage 2, and set the scene for eventual full utilization of the country's abundant thorium resources as fuel for nuclear reactors.

Japan plans to develop FBRs, and its Joyo experimental reactor which has been operating since 1977 is now being boosted to 140 MWt. The 280 MWe Monju prototype commercial FBR was connected to the grid in 1995, but was subsequently shut down due to a sodium leak.

The Russian BN-600 fast breeder reactor has been supplying electricity to the grid since 1981 and has the best operating and production record of all Russia's nuclear power units. It uses a uranium oxide fuel and the sodium coolant delivers 550°C at little more than atmospheric pressure. The BN-350 FBR operated in Kazakhstan for 27 years and about half of its output was used for water desalination. Russia plans to reconfigure the BN-600 to burn plutonium from its military stockpiles.

Construction has started at Beloyarsk on the first BN-800, a new larger (880 MWe) FBR from OKBM with improved features including fuel flexibility—U+Pu nitride, MOX, or metal—and with breeding ratio up to 1.3. It has much enhanced safety and improved economy with operating costs expected to be only 15% more relative to VVER. It is capable of burning 2 tonnes of plutonium per year from dismantled weapons and will test the recycling of minor actinides in the fuel. However, industry spokesmen have warned the government that Russia's world leadership in FBR development is threatened due to lack of funding for completion of BN-800.

Russia has experimented with several lead-cooled reactor designs, and has used lead-bismuth cooling for 40 years in nuclear reactors for its 7 Alfa class submarines. Lead-208 (²⁰⁸Pb) (54% of naturally-occurring lead) is transparent to neutrons. A significant new Russian design is the BREST fast neutron reactor, of 300 MWe or more with lead as the primary coolant, at 540°C, and with supercritical steam generators. It is inherently safe and uses a high-density U+Pu nitride fuel with no requirement for high enrichment levels. No weapons-grade plutonium can be produced—since there is no uranium blanket, all breeding occurs in the core. The initial cores can comprise Pu and spent fuel. Subsequently, any surplus plutonium can be used in the cores of new reactors. Spent fuel can be recycled indefinitely, with on-site reprocessing and associated facilities. A pilot unit is being built at Beloyarsk and 1200 MWe units are planned.

In the USA, GE was involved in designing a modular 150 MWe liquid metal-cooled inherently-safe reactor—PRISM. GE and Argonne National Laboratory (ANL) have also been developing an advanced liquid-metal fast breeder reactor (ALMR) of over 1400 MWe, but both designs at an early stage were withdrawn from NRC review. No US fast neutron reactor has so far been larger than 66 MWe and none has supplied electricity commercially.

The Super-PRISM is a GE advanced reactor design for compact modular pool-type nuclear reactors with passive cooling and radioactive decay heat removal. Modules are 1000 MWt and operate at a higher temperature—510°C—than the original PRISM. The pool-type modules contain the complete primary system with sodium coolant. The plutonium (Pu) and depleted uranium (DU) fuel can be oxide or metal, but minor actinides are not removed in reprocessing so that even fresh fuel is intensely radioactive and hence resistant to misappropriation. The fission products are removed in reprocessing and resultant wastes are shorter-lived than usual. Fuel stays in the reactor six years, with one-third removed every two years. The commercial plant concept uses six reactor modules to provide 2280 MWe, and the design meets Generation IV criteria including generation cost of under 3 cents/kWh.

Accelerator-Driven Systems

A recent development has been the merging of accelerator and fission reactor technologies to generate electricity and transmute long-lived radioactive wastes.

A high-energy proton beam hitting a heavy metal target produces neutrons by spallation. The neutrons cause fission in the fuel, but unlike in a conventional nuclear reactor, the fuel is sub-critical, and fission ceases when the accelerator is turned off. The fuel may be uranium, plutonium or thorium, possibly mixed with long-lived wastes from conventional reactors.

Many technical and engineering questions remain to be explored before the potential of this concept can be demonstrated.

Further Reading

• Hore-Lacy, Ian (Lead Author); World Nuclear Association (Content Partner); Cutler J. Cleveland (Topic Editor). 2008. Advanced nuclear power reactors. In: Encyclopedia of Earth. Eds. Cutler J. Cleveland (Washington, D.C.: Environmental Information Coalition, National Council for Science and the Environment). [First published in the Encyclopedia of Earth August 26, 2006; Last revised February 12, 2008; Retrieved October 30, 2008].