Advanced Nuclear Power Reactors

(April 2008)

The next two generations of nuclear reactors are currently being developed in several countries.
The first (3rd generation) advanced reactors have been operating in Japan since 1996. Late 3rd generation designs are now being built.
Newer advanced reactors have simpler designs which reduce capital cost. They are more fuel efficient and are inherently safer.

The nuclear power industry has been developing and improving reactor technology for more than five decades and is starting to build the next generations of reactors to fill orders now materialising.
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 (and 3+) 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 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 These and other second-generation nuclear power units have been found to be safe and reliable, but they are being superseded by better designs.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 standardised 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 the amount of waste,
burnable absorbers ("poisons") to extend fuel life.

The greatest departure from second-generation designs is that many incorporate passive or inherent safety features* which 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 reactor safety systems are 'active' in the sense that they involve electrical or mechanical operation on command. Some engineered systems operate passively, eg pressure relief valves. They function without operator control and despite any loss of auxiliary power. 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 predecessors. Increasingly they involve international collaboration.
Certification of designs is on a national basis, and is safety-based. In Europe there are moves towards harmonised requirements for licensing.
However, in Europe reactors may also be certified according to compliance with European Utilities Requirements (EUR). These are basically a utilities' wish list of some 5000 items needed for new nuclear plants. Plants certified as complying with EUR include Westinghouse AP1000, Gidropress' AES-92, Areva's EPR, GE's ABWR, Areva's SWR-1000, and Westinghouse BWR 90.

In the USA a number of reactor types have received Design Certification (see below) and others are in process: ESBWR from GE-Hitachi, US EPR from Areva and US-APWR from Mitsubishi. Early in 2008 the NRC said that beyond these three, six pre-application reviews would get underway until about 2010. These include IRIS from Westinghouse, PBMR from Eskom and 4S from Toshiba as well as Canada's ACR and General Atomics' GT-MHR apparently. See also NRC.

Longer term, NRC expected to focus on the Next-Generation Nuclear Plant (NGNP) for the USA (see USA paper) - essentially the Very High Temperature Reactor (VHTR) among the >Generation IV designs.

Joint Initiatives

Two major international initiatives have been launched to define future reactor and 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 which has identified six reactor concepts for further investigation with a view to commercial deployment by 2030. See Generation IV paper and DOE web site on "4th generation reactors".

The 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.

At the commercial level, by the end of 2006 three major Western-Japanese alliances had formed to dominate much of the world reactor supply market:

Areva with Mitsubishi Heavy Industries (MHI) in a major project and subsequently in fuel fabrication,
General Electric with Hitachi as a close relationship
Westinghouse had become a 77% owned subsidiary of Toshiba (with Shaw group 20%).

Then in March 2008 Toshiba signed a technical cooperation agreement on civil nuclear power with Russia's Atomenergoprom - the single vertically-integrated state holding company for Russia's nuclear power sector created in 2007. This could lead to a "strategic partnership" and include designing and engineering of commercial nuclear power plants, as well as manufacturing and maintenance of large equipment.

Light Water Reactors

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) derived from a General Electric design. Two examples built by Hitachi and two by Toshiba are in commercial operation in Japan, with another under construction there and two in Taiwan. Four more are planned in Japan and another two in the USA. Though GE and Hitachi have subsequently joined up, Toshiba retains some rights over the design. Both GE-Hitachi and Toshiba (with NRG Energy in USA) are marketing the design.

The other type, System 80+, is an advanced pressurised water reactor (PWR), which was ready for commercialisation 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.

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. US utilities will be able to obtain a single NRC licence to both construct and operate a reactor before construction begins.

Separate from the NRC process and beyond its immediate requirements, the US nuclear industry selected one standardised 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, was half funded by DOE. It means that prospective buyers now have firm information on construction costs and schedules.

The Westinghouse AP-1000, scaled-up from the AP-600, received final design certification from the NRC in December 2005 - the first generation 3+ type to do so. It represents the culmination of a 1300 man-year and $440 million design and testing program. In May 2007 Westinghouse applied for UK generic design assessment (pre-licensing approval) based on the NRC design certification, and expressing its policy of global standardisation. The application was supported by utilities including E.ON.

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 its has a 60 year operating life. It has been selected for building in China (4 units) and is under active consideration for building in Europe and 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. Originally the European Simplified Boiling Water Reactor, this is now known as the Economic & Simplified BWR (ESBWR) and a 1560 MWe version is at preliminary stage of NRC design certification in the USA, so that design approval is expected at the end of 2008, with formal certification 12 months later. It is favoured for early US construction and could be operational in 2014. It uses 4.2% enriched fuel and has a design life of 60 years.

Another US-origin but international project which is a few years behind the AP-1000 is the International Reactor Innovative & Secure (IRIS). Westinghouse is leading a wide consortium developing it as an advanced 3rd Generation project. IRIS is a modular 335 MWe pressurised water reactor with integral steam generators and primary coolant system all within the pressure vessel. It is nominally 335 MWe but can be less, eg 100 MWe. Fuel is initially similar to present LWRs with 5% enrichment and burn-up of 60,000 MWd/t with fuelling interval of 3 to 3.5 years, but is designed ultimately for 10% enrichment and 80 GWd/t burn-up with an 8 year cycle, or equivalent MOX core. The core has low power density. IRIS could be deployed in the next decade, and US design certification is at pre-application stage. Multiple modules are expected to cost US$ 1000-1200 per kW for power generation, though some consortium partners are interested in desalination, one in district heating.

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 GE-Hitachi-Toshiba units cost about US$ 2000/kW to build, and produce power at about US 7c/kWh. Two more started up in 2004 & 2005. Future ABWR units are expected to cost US$ 1700/kW. Several of the 1350 MWe units are under construction and planned in Japan and Taiwan.

To complement this ABWR Hitachi-GE 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 standardised features which reduce costs. Construction of the ABWR-600 is expected to take 34 months - significantly less than the 1350 MWe units.

Mitsubishi's large APWR (1538 MWe) - advanced PWR - was developed in collaboration with four utilities (Westinghouse was earlier involved). The first two are planned for Tsuruga. It is simpler, combines active and passive cooling systems to greater effect, and has over 55 GWd/t fuel burn-up. It will be the basis for the next generation of Japanese PWRs.

The US-APWR will be 1700 MWe, due to higher thermal efficiency (39%) and has 24 month refuelling cycle and target cost of $1500/kW. US design certification application was in January 2008 with approval expected in 2011. The first units may be built for TXU at Comanche Peak near Dallas, Texas. In March 2008 MHI submitted the same design for EUR certification, as EU-APWR.

The Atmea joint venture has been established by Areva NP and Mitsubishi Heavy Industries to develop an 1100 MWe (net) three-loop PWR with extended fuel cycles, 37% thermal efficiency and the capacity to use mixed-oxide fuel only. Fuel cycle is 12-24 months and the reactor has load-following capability. They expect to have this ready for licence application by 2010. The reactor is regarded as mid-sized relative to other generation III units and will be marketed primarily to countries embarking upon nuclear power programs.

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 about 2012. 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 60 years.

In Europe, several designs are being developed to meet the European Utility Requirements (EUR) of French and German utilities, which have stringent safety criteria. Areva NP (formerly Framatome ANP) has developed a large (1600 and up to 1750 MWe) European pressurised 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%. It is capable of using a full core load of MOX. Availability is expected to be 92% over a 60-year service life. It has four separate, redundant safety systems rather than passive safety.

The first EPR unit is being built at Olkiluoto in Finland, the second at Flamanville in France. A US version, the US-EPR, is undergoing review in USA with intention of a design certification application in 2007. It is now known as the Evolutionary PWR (EPR). Overnight capital cost is quoted as $2400 per kilowatt, levelised over the first four units.

Together with German utilities and safety authorities, Areva NP (Framatome ANP) is also developing another evolutionary design, the SWR 1000, a 1200-1290 MWe BWR with 60 year design life. The design was completed in 1999 and US certification was sought, but then deferred. As well as many passive safety features, the reactor is simpler overall and uses high-burnup fuels enriched to 3.54%, giving it refuelling intervals of up to 24 months. It is ready for commercial deployment and the prospects of that will be helped by a 2008 agreement with Siemens and the major German utility E.On (Siemens built the Gundremmingen plant on which the design is based, for E.On).
Toshiba has been developing its evolutionary advanced BWR (1500 MWe) design, originally BWR 90+ from ABB then Westinghouse, working with Scandinavian utilities to meet EUR requirements.

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

Gidropress late-model VVER-1000 units with enhanced safety (AES 92 & 91 power plants) are being built in India and China. Two more are planned for Belene in Bulgaria. The AES-92 is certified as meeting EUR.

A third-generation standardised VVER-1200 reactor of 1150-1200 MWe is an evolutionary development of the well-proven VVER-1000 in the AES-92 plant, with longer life, greater power and efficiency. The lead units will be built at Novovoronezh II, to start operation in 2012-13 followed by Leningrad II for 2013-14. An AES-2006 plant will consist of two of these OKB Gidropress reactor units expected to run for 50 years with capacity factor of 90%. Capital cost is said to be US$ 1200/kW and construction time 54 months. They have enhanced safety including that related to earthquakes and aircraft impact with some passive safety features, double containment and core damage frequency of 1x10^-7.

Atomenergoproekt say that the AES-2006 conforms to both Russian standards and European Utilities Requirements (EUR).

The VVER-1500 model was being developed by Gidropress. It will have 50-60 MWd/t burn-up and enhanced safety. Design was expected to be complete in 2007 but this schedule has slipped in favour of the evolutionary VVER-1200.

OKBM's VBER-300 PWR is a 295-325 MWe unit developed from naval power plants and was originally envisaged in pairs as a floating nuclear power plant. It is designed for 60 year life and 90% capacity factor. It now planned to develop it as a land-based unit with Kazatomprom, with a view to exports, and the first unit will be built in Kazakhstan.

The VBER-300 and the similar-sized VK300 are more fully described in the Small Nuclear Power Reactors paper.

Heavy Water Reactors

Canada has had two designs under development which are based on its reliable CANDU-6 reactors, the most recent of which are operating in China.

The CANDU-9 (925-1300 MWe) was developed from this also as a single-unit plant. It has flexible fuel requirements ranging from natural uranium through slightly-enriched uranium, recovered uranium from reprocessing spent PWR fuel, mixed oxide (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 early in 1997, but the design has been shelved.

Some of the innovation of this, along with experience in building recent Korean and Chinese units, was then put back into the Enhanced CANDU-6 - built as twin units - with power increase to 750 MWe and flexible fuel options, plus 4.5 year construction and 60-year plant life (with mid-life pressure tube replacement). This is under consideration for new build in Ontario.

The Advanced Candu Reactor (ACR), a 3rd generation reactor, is a more innovative concept. While retaining the low-pressure heavy water moderator, it incorporates some features of the pressurised 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 design was 700 MWe but is physically much smaller, simpler and more efficient as well as 40% cheaper than the CANDU-6. But the ACR-1000 of 1080-1200 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). The ACR will run on low-enriched uranium (about 1.5-2.0% U-235) with high burn-up, extending the fuel life by about three times and reducing high-level waste volumes accordingly. It will also efficiently burn MOX fuel, thorium and actinides.

Regulatory confidence in safety is enhanced by a small negative void reactivity for the first time in CANDU, and utilising other passive safety features as well as two independent and fast shutdown systems. Units will be assembled from prefabricated modules, cutting construction time to 3.5 years. ACR units can be built singly but are optimal in pairs. They will have 60 year design life overall but require mid-life pressure tube replacement.

ACR is moving towards design certification in Canada, with a view to following in China, USA and UK. In 2007 AECL applied for UK generic design assessment (pre-licensing approval). The first ACR-1000 unit is expected to be operating in 2016 in Ontario.

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

India is developing the Advanced Heavy Water reactor (AHWR) as the third stage in its plan to utilise thorium to fuel its overall nuclear power program. The AHWR is a 300 MWe 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 Th-U-233 oxide pins and 24 Pu-Th 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 U-233 bred from Th-232 and have a low Pu inventory and consumption, with slightly negative void coefficient of reactivity. It is designed for 100 year plant life and is expected to utilise 65% of the energy of the fuel.

Once it is fully operational, each AHWR fuel assembly will have the fuel pins arranged in three concentric rings arranged:

Inner: 12 pins Th-U-233 with 3.0% U-233,
Intermediate: 18 pins Th-U-233 with 3.75% U-233,
Outer: 24 pins Th-Pu-239 with 3.25% Pu.

The fissile plutonium content will decrease from an initial 75% to 25% at equilibrium discharge burn-up level.

High-Temperature Gas-Cooled Reactors

These reactors use helium as a coolant which at up to 950ºC drives a gas turbine for electricity and a compressor to return the gas to the reactor core. Fuel is in the form of TRISO particles less than a millimetre in diameter. Each has a kernel of uranium oxycarbide, with the uranium enriched up to 17% U-235. 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 - 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 a step change in safety, economics and proliferation resistance. Production units will be 165 MWe. They will have a direct-cycle gas turbine generator and thermal efficiency 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 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 central 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 reactor shuts itself down - giving inherent safety. Each unit will finally discharge about 19 tonnes/yr of spent pebbles to ventilated on-site storage bins.
Overnight construction cost (when in clusters of eight units) is expected to be US$ 1000/kW and generating cost below 3 US 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 2007 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 the core is replaced every 18 months. Burn-up is about 100,000 MWd/t. It 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.

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 in reactor fuel as well as the fissile U-235 isotope used in most 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 % U-235 and 99.3 % U-238. In any reactor the U-238 component is turned into several isotopes of plutonium during its operation. Two of these, Pu 239 and Pu 241, then undergo fission in the same way as U 235 to produce heat. In a fast neutron reactor this process is optimised so that it can 'breed' fuel, often using a depleted uranium blanket around the core. FBRs can utilise 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 there is employed to explore the use of thorium as nuclear fuel, by breeding fissile U-233. 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 U-233. This will take India's ambitious thorium program to stage 2, and set the scene for eventual full utilisation of the country's abundant thorium to fuel 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 then 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 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 the 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 - operating cost is expected to be only 15% more than 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.

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 and as of 2006 funding seems to have been released.

Russia has experimented with several lead-cooled reactor designs, and has used lead-bismuth cooling for 40 years in reactors for its 7 Alfa class submarines. Pb-208 (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 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 the breeding occurs in the core). The initial cores can comprise Pu and spent fuel - hence loaded with fission products, and radiologically 'hot'. Subsequently, any surplus plutonium, which is not in pure form, can be used as the cores of new reactors. Used fuel can be recycled indefinitely, with on-site reprocessing and associated facilities. A pilot unit is planned for Beloyarsk and 1200 MWe units are proposed.

In the USA, GE was involved in designing a modular 150 MWe liquid metal-cooled inherently-safe reactor - PRISM. GE and Argonne 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 reactors with passive cooling and decay heat removal. Modules are 1000 MWt and operate at higher temperature - 510C, than the original PRISM. The pool-type modules contain the complete primary system with sodium coolant. The Pu & 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.

Korea's KALIMER (Korea Advanced LIquid MEtal Reactor) is a 600 MWe pool type sodium-cooled fast reactor designed to operate at over 500ºC. It has evolved from a 150 MWe version. It has a transmuter core, and no breeding blanket is involved. Future development of KALIMER as a Generation IV type is envisaged.

In the USA Mitsubishi Heavy Industries (MHI) is involved with a consortium to develop the Advanced Recycling Reactor, a fast reactor which will burn actinides with uranium and plutonium. This will be based on MHI's Japan Standard Fast reactor concept, though with breeding ration less than 1:1. In this connection MHI has also set up Mitsubishi FBR Systems (MFBR).

See also paper on Fast Neutron Reactors.

Generation IV Designs

See paper on six Generation IV Reactors, also DOE paper.

Small Reactors

See also paper on Small Nuclear Power Reactors for other advanced designs, mostly under 300 MWe.

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 a conventional 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. See also ADS briefing paper.

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Lauret, P. et al, 2001, The Nuclear Engineer 42, 5.
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AECL Candu-6 & ACR publicity, late 2005.

Advanced Thermal Reactors being marketed

Country and developer	Reactor	Size MWe	Design Progress	Main Features (improved safety in all)
US-Japan (GE-Hitachi, Toshiba)	ABWR	1300	Commercial operation in Japan since 1996-7. In US: NRC certified 1997, FOAKE.	Evolutionary design. More efficient, less waste. Simplified construction (48 months) and operation.
USA (Westinghouse)	AP-600 AP-1000 (PWR)	600 1100	AP-600: NRC certified 1999, FOAKE. AP-1000 NRC certification 2005.	Simplified construction and operation. 3 years to build. 60-year plant life.
France-Germany (Areva NP)	EPR US-EPR (PWR)	1600	Future French standard. French design approval. Being built in Finland. US version developed.	Evolutionary design. High fuel efficiency. Low cost electricity.
USA (GE)	ESBWR	1550	Developed from ABWR, under certification in USA	Evolutionary design. Short construction time.
Japan (utilities, Mitsubishi)	APWR US-APWR EU-APWR	1530 1700 1700	Basic design in progress, planned for Tsuruga US design certification application 2008.	Hybrid safety features. Simplified Construction and operation.
South Korea (KHNP, derived from Westinghouse)	APR-1400 (PWR)	1450	Design certification 2003, First units expected to be operating c 2012.	Evolutionary design. Increased reliability. Simplified construction and operation.
Germany (Areva NP)	SWR-1000 (BWR)	1200	Under development, pre-certification in USA	Innovative design. High fuel efficiency.
Russia (Gidropress)	VVER-1200 (PWR)	1200	Replacement for Leningrad and Novovoronezh plants	High fuel efficiency.
Russia (Gidropress)	V-392 (PWR)	950-1000	Two being built in India, Bid for China in 2005.	Evolutionary design. 60-year plant life.
Canada (AECL)	CANDU-6 CANDU-9	750 925+	Enhanced model Licensing approval 1997	Evolutionary design. Flexible fuel requirements. C-9: Single stand-alone unit.
Canada (AECL)	ACR	700 1080	undergoing certification in Canada	Evolutionary design. Light water cooling. Low-enriched fuel.
South Africa (Eskom, Westinghouse)	PBMR	170 (module)	prototype due to start building (Chinese 200 MWe counterpart under const.)	Modular plant, low cost. High fuel efficiency. Direct cycle gas turbine.
USA-Russia et al (General Atomics - OKBM)	GT-MHR	285 (module)	Under development in Russia by multinational joint venture	Modular plant, low cost. High fuel efficiency. Direct cycle gas turbine.