Advanced Nuclear Power Reactors
(Updated April 2013)
- Improved designs of nuclear power reactors are currently being developed in several countries.
- The first so-called 3rd generation advanced reactors have been operating in Japan since 1996.
- Newer advanced reactors now being built 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 generation of nuclear power reactors to fill new orders.
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 and French fleets and most in operation elsewhere. So-called Generation III (and 3+) are the Advanced Reactors discussed in this paper, though the distinction from Generation II is arbitrary. 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 These and other nuclear power units now operating have been found to be safe and reliable, but they are being superseded by better designs.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 elsewhere 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.
So-called 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,
- further reduced possibility of core melt accidents,*
- substantial grace period, so that following shutdown the plant requires no active intervention for (typically) 72 hours,
- resistance to serious damage that would allow radiological release from an aircraft impact,
- higher burn-up to use fuel more fully and efficiently and reduce the amount of waste,
- greater use of burnable absorbers ("poisons") to extend fuel life.
* The US NRC requirement for calculated core damage frequency is 1x10-4, most current US plants have about 5x10-5 and Generation III plants are about ten times better than this. The IAEA safety target for future plants is 1x10-5. Calculated large release frequency (for radioactivity) is generally about ten times less than CDF.
The greatest departure from most designs now in operation 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, but these terms are not properly used to characterise whole reactors.
Another departure is that some PWR types will be designed for load-following. While most French reactors today are operated in that mode to some extent, the EPR design has better capabilities. It will be able to maintain its output at 25% and then ramp up to full output at a rate of 2.5% of rated power per minute up to 60% output and at 5% of rated output per minute up to full rated power. This means that potentially the unit can change its output from 25% to 100% in less than 30 minutes, though this may be at some expense of wear and tear.
Many are larger than predecessors. Increasingly they involve international collaboration.
However, certification of designs is on a national basis, and is safety-based. In Europe there are moves towards harmonised requirements for licensing. In Europe, reactors may also be certified according to compliance with European Utilities Requirements (EUR) of 12 generating companies, which have stringent safety criteria. The EUR 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 Kerena, and Westinghouse BWR 90.
European regulators are increasingly requiring large new reactors to have some kind of core catcher or similar device, so that in a full core-melt accident there is enhanced provision for cooling the bottom of the reactor pressure vessel or simply catching any material that might melt through it. The EPR and VVER-1200 have core-catchers under the pressure vessel, the AP1000 and APWR have provision for enhanced water cooling.
The UK’s Office for Nuclear Regulation (ONR) is processing the Areva EPR and Westinghouse AP1000 designs for Generic Design Assessment (GDA). The initial assessment for these was completed in mid 2011. The ONR and Environment Agency jointly issued interim design acceptance confirmations (iDAC), and interim statements on design acceptability (iSODA) for the two designs in mid December 2011. A full DAC and SODA may be issued for the UK EPR by the end of 2012, but Westinghouse decided to request a pause in the GDA process pending customer input to finalizing it.
As the GDA has proceeded issues have arisen which are in common with new capacity being built elsewhere, particularly the EPR units in Finland and France. This has led to international collaboration and a joint regulatory statement on the EPR control and instrumentation among ONR, US NRC, France's ASN and Finland's STUK. More broadly it relates to the Multinational Design Evaluation Program and will help improve the harmonization of regulatory requirements internationally.
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. The ESBWR is on track to receive certification in 2013, and the US EPR in 2015, and the US-APWR early in 2016. An application for Korea’s APR1400 was expected in mid 2013.
Longer term, the NRC expected to focus on the Next-Generation Nuclear Plant (NGNP) for the USA (see US Nuclear Power Policy paper ) - essentially the Very High Temperature Reactor (VHTR) among the Generation IV designs.
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:
* GEH is the main international partnership, 60% GE. In Japan it is Hitachi GE, 80% owned by Hitachi.
Subsequently there have been a number of other international collaborative arrangements initiated among reactor vendors and designers, but it remains to be seen which will be most significant.
US Design certification
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 and now promoted both by GE-Hitachi and Toshiba as a proven design, which is in service.
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 from 2013 and is being 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 utility requirements for advanced reactors. Both GE Hitachi and Toshiba in 2010 submitted separate applications to renew the design certification for their respective versions of ABWR (Toshiba's incorporating design changes submitted to NRC already in connection with application for the South Texas Project). The Japanese version of it differs in allowing modular construction, so is not identical to that licenced in the USA.
Another, more innovative US advanced reactor is smaller - 600 MWe - and has passive safety features (its projected core damage frequency is more than 100 times less than today's NRC requirements). The Westinghouse AP600 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 AP600, for detailed first-of-a-kind engineering (FOAKE) work. The US$ 200 million program was half funded by DOE and means that prospective buyers now have fuller information on construction costs and schedules.
The 1100 MWe-class Westinghouse AP1000, scaled-up from the AP600, received final design certification from the NRC in December 2005 - the first Generation 3+ type to do so. It represented 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 European utilities.
Overnight capital costs were originally projected at $1200 per kilowatt and modular design is expected to reduce construction time eventually to 36 months. The AP1000 generating costs are also expected to be very competitive and it has a 60-year operating life. It is being built in China (4 units under construction, with many more to follow) and is under active consideration for building in Europe and USA. It is capable of running on a full MOX core if required.
In February 2008 the NRC accepted an application from Westinghouse to amend the AP1000 design, and this review was completed with revised design certification in December 2011. The NRC chairman said that the revised AP1000 design is one that seems to most fully meet the expectations of the commission’s policy statement on advanced reactors.” “The design provides enhanced safety margins through use of simplified, inherent, passive or other innovative safety and security functions, and also has been assessed to ensure it could withstand damage from an aircraft impact without significant release of radioactive materials."
A contrast between the 1188 MWe Westinghouse reactor at Sizewell B in the UK and the modern AP1000 of similar-power illustrates the evolution from 1970-80 types. First, the AP1000 footprint is very much smaller - about one quarter the size, secondly the concrete and steel requirements are less by a factor of five*, and thirdly it has modular construction. A single unit will have 149 structural modules of five kinds, and 198 mechanical modules of four kinds: equipment, piping & valve, commodity, and standard service modules. These comprise one third of all construction and can be built off site in parallel with the on-site construction.
*Sizewell B: 520,000 m3 concrete (438 m3/MWe), 65,000 t rebar (55 t/MWe);
AP1000: <100,000 m3 concrete (90 m3/MWe, <12,000 t rebar (11 t/MWe).
At Sanmen in China, where the first AP1000 units are under construction, the first module lifted into place weighed 840 tonnes. More than 50 other modules used in the reactors' construction weigh more than 100 tonnes, while 18 weigh in excess of 500 tonnes.
Light Water Reactors
(power reactors moderated and cooled by water)
Areva NP (formerly Framatome ANP) has developed a large (4590 MWt, typically 1750 MWe gross and 1630 MWe net) 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 a 4-loop design derived from the German Konvoi types with features from the French N4, 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 a high thermal efficiency, of 37%, and net efficiency of 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 double containment with four separate, redundant active safety systems, and boasts a core catcher under the pressure vessel. The safety systems are physically separated through four ancillary buildings on the same concrete raft, and two of them are aircraft crash protected. The primary diesel generators have fuel for 72 hours, the secondary back-up ones for 24 hours, and tertiary battery back-up lasts 12 hours. It is designed to withstand seismic ground acceleration of 600 Gal without safety impairment.
The first EPR unit is being built at Olkiluoto in Finland, the second at Flamanville in France, the third European one will be at Penly in France, and two further units are under construction at Taishan in China.
A US version, the US-EPR quoted as 1710 MWe gross and about 1580 MWe net, was submitted for US design certification in December 2007, and this is expected to be granted in mid 2014. The first unit (with 80% US content) was expected to be grid connected by 2020. It is now known as the Evolutionary PWR (EPR). Much of the one million man-hours of work involved in developing this US EPR is making the necessary changes to output electricity at 60 Hz instead of the original design's 50 Hz. The main development of the type was to be through UniStar Nuclear Energy.
The Westinghouse AP1000 is a 2-loop PWR which has
evolved from the smaller AP600, one of the first new reactor designs certified
by the US NRC, in 2005. Simplification was a major design objective of the
AP1000, in overall safety systems, normal operating systems, the control room,
construction techniques, and instrumentation and control systems provide cost
savings with improved safety margins. It has a core cooling system including
passive residual heat removal by convection, improved containment isolation,
passive containment cooling system to the atmosphere and in-vessel retention of
core damage (corium) with water cooling around it. No safety-related pumps or
ventilation systems are needed. It is being built in China, and the Vogtle site
is being prepared for initial units in USA. The first four units are on
schedule, being assembled from modules. It is quoted as 1200 MWe gross and 1117
MWe net (3400 MWt), though 1250 MWe gross in China. Westinghouse earlier claimed
a 36 month construction time to fuel loading. The first ones being built in
China are on a 51-month timeline to fuel loading, or 57-month schedule to grid
connection. After the first four units there, the design is known as CAP1000.
Westinghouse is working with SNPTC and SNERDI in China
to develop jointly a passively safe 1520 MWe (4040 MWt) 2-loop design from the
AP1000, the CAP1400, with 193
fuel assemblies and improved steam generators, operating at 304°C, 60-year design life, and 72-hour
non-intervention period in event of accident. This may extend to a larger,
3-loop CAP1700 or CAP 2100 design if the passive cooling system can be scaled to
that level. Westinghouse has agreed that SNPTC will own the intellectual
property rights for any AP1000 derivatives over 1350 MWe.
The advanced boiling water reactor (ABWR) is derived from a General Electric design in collaboration with Toshiba. Two examples built by Hitachi and two by Toshiba are in commercial operation in Japan (1315 MWe net), with another two under construction there and two in Taiwan. Four more are planned in Japan and another two in the USA.
The ABWR is now offered in slightly different versions by GE Hitachi, Hitachi-GE and Toshiba, so that 'ABWR" is now a generic term. It is basically a 1380 MWe (gross) unit (3926 MWt in Toshiba version), though GE Hitachi quote 1350-1600 MWe net. Toshiba outlines development from its 1400 MWe class to a 1500-1600 MWe class unit (4300 MWt). Tepco was funding the design of a next generation BWR, and the ABWR-II is quoted as 1717 MWe.
Toshiba is promoting its EU-ABWR of 1600 MWe with core catcher and filtered vent, developed with Westinghouse Sweden. The Hitachi UK-ABWR may have similar features but be similar size to Japanese units.
The first four ABWRs were each built in 39-43 months on a single-shift basis. Though GE and Hitachi have subsequently joined up, Toshiba retains some rights over the design, as does Tepco. The design can run on full-core mixed oxide (MOX) fuel, as for the Ohma plant being built in Japan. Design life is 60 years. It has a high level of active safety. Unlike previous BWRs in Japan the external recirculation loop and internal jet pumps are replaced by coolant pumps mounted at the bottom of the reactor pressure vessel. Safety systems are active
Both Toshiba and GE-Hitachi have applied separately to NRC for design certification renewal. The initial certification in 1997 was for 15 years and in 2011 the NRC certified for GE Hitachi an evolved version which allows for aircraft impacts. Hitachi has applied for UK Generic Design Assessment for its version of the ABWR.
GE Hitachi Nuclear Energy's ESBWR is an improved design that utilizes passive safety features and natural circulation principles and is essentially an evolution from a predecessor design, the SBWR at 670 MWe. GE-H says it is safer and more efficient than earlier models, with 25% fewer pumps, valves and motors, and can maintain cooling for six days after shutdown with no AC or battery power. The emergency core cooling system has eliminated the need for pumps, using passive and stored energy.
The ESBWR (4500 MWt) will produce approximately 1600 MWe gross, and 1535 MWe net, depending on site conditions, and has a design life of 60 years. It was more fully known as the Economic & Simplified BWR (ESBWR) and leverages proven technologies from the ABWR. The ESBWR is in advanced stages of licensing review with the US NRC for GE Hitachi and is on schedule for full design certification in 2013. Design approval was in March 2011. It was submitted for UK Generic Design Assessment in 2007, but a year later GE-H requested that this be suspended.
GEH is selling this alongside the ABWR, which it characterises as more expensive to build and operate, but proven. ESBWR is more innovative, with lower building and operating costs and a 60-year life.
Mitsubishi's large APWR - advanced PWR of 1538 MWe gross (4451 MWt) - was developed in collaboration with four utilities (Westinghouse was earlier involved). The first two are planned for Tsuruga, coming on line from 2016. It is a 4-loop design with 257 fuel assemblies and neutron reflector, is simpler, combines active and passive cooling systems in double containment, and has over 55 GWd/t (and up to 62 GWd/t) fuel burn-up. It is the basis for the next generation of Japanese PWRs. The planned APWR+ is 1750 MWe and has full-core MOX capability.
The US-APWR will be 1700 MWe gross, about 1620 MWe net, due to longer (4.3m) fuel assemblies, higher thermal efficiency (37%) and has 24 month refuelling cycle. Its emergency core cooling system (ECCS) has four independent trains, and its outer walls and roof are 1.8 m thick. US design certification application was in January 2008 with certification expected in 2016. In March 2008 MHI submitted the same design for EUR certification, as EU-APWR, and it will join with Iberdrola Engineering & Construction in bidding for sales of this in Europe. Iberdrola would be responsible for building the plants.
The Japanese government is expected to provide financial support for US licensing of both US-APWR and the ESBWR. The Washington Group International will be involved in US developments with Mitsubishi Heavy Industries (MHI). The US-APWR has been selected by Luminant for Comanche Peak, Texas, and when the COL application for the new reactors was lodged Luminant and MHI announced a joint venture to build and own the twin-unit plant. This Comanche Peak Nuclear Power Co is 88% Luminant, 12% MHI.
South Korea's APR1400 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. It is 1455 MWe gross in Korean conditions according to IAEA status report, 1350-1400 MWe net (3983 - nominal 4000 MWt) with 2-loop primary circuit. The first of these is under construction - Shin-Kori-3 & 4, expected to be operating in 2013. Fuel has burnable poison and will have up to 55 GWd/t burn-up, refueling cycle c 18 months, outlet temperature 324ºC. Projected cost at the end of 2009 was US$ 2300 per kilowatt, with 48-month construction time. Plant life is 60 years, seismic design basis is 300 Gal. A low-speed (1800 rpm) turbine is envisaged. It has been chosen as the basis of the United Arab Emirates nuclear program on the basis of cost and reliable building schedule, and is being offered in Finland. Following pre-application meetings with NRC since 2010, an application for US Design Certification is planned in mid 2013.
Based on this KOPEC is developing an EU version (APR1400-EUR) with double containment and core-catcher, and a more advanced 1550 MWe (gross) version, the APR+.
In addition some of the APR features are being incorporated into a development of the OPR-1000 to give an exportable APR-1000 intended for overseas markets, notably Middle East and Southeast Asia, and will be able to operate with an ultimate heat sink of 40°C, instead of 35°C for the OPR-1000. Improved safety and performance will raise the capital cost above that of the OPR, but it this will be offset by reduced construction time (40 months instead of 46) due to modular construction.
The Atmea 1 has been developed by the Atmea joint venture established in 2007 by Areva NP and Mitsubishi Heavy Industries to produce an evolutionary 1150 MWe net 93150 MWt) three-loop PWR using the same steam generators as EPR. This has 37% net thermal efficiency, 157 fuel assemblies 4.2 m long, 60-year life, and the capacity to use mixed-oxide fuel only. Fuel cycle is flexible 12 to 24 months with short refuelling outage and the reactor has load-following and frequency control capability.
Following an 18-month review, the French regulator ASN approved the general design in February 2012. The reactor is regarded as mid-sized relative to other modern designs and will be marketed primarily to countries embarking upon nuclear power programs. It has three active and passive redundant safety systems and an additional backup cooling chain, similar to EPR. Canadian vendor pre-project design review is under way.
Early in 2012 Areva NP and EdF agreed in principle with China Guangdong Nuclear Power group to develop a mid-size PWR on the basis of CGNPC’s CPR1000, with third-generation safety features. A further 3-way agreement was signed in September, with a view to having an outcome by mid 2013. It is not clear whether Mitsubishi Heavy Industries might be involved, though Areva has said that it wants the design "to have the highest possible technical convergence" with Atmea1. If a new reactor design results, it would be a competitor for Atmea1. However, Areva says that the talks are not aimed at joint development of a 1000 MWe reactor, so much as "to see if the three companies can converge on specifications for such a design that would allow deeper collaboration".
Together with German utilities and safety authorities, Areva NP has also developed another evolutionary design, the Kerena, a 1290 MWe gross, 1250 MWe net (3370 MWt) BWR with 60-year design life formerly known as SWR 1000, The design, based on the Gundremmingen plant built by Siemens, was completed in 1999 and US certification was sought, but then deferred. It has not yet been submitted for certification anywhere.
It has two redundant active safety systems and two passive safety systems, including a core-catcher, similar to EPR. The reactor is simpler overall and uses high-burnup fuels enriched to 3.54%, giving it refuelling intervals of up to 24 months. It has 37% net efficiency and is ready for commercial deployment.
Gidropress late-model VVER-1000 units with enhanced safety (AES 92 & 91 power plants) are being built in India and China. Two more (V466B variant) were planned for Belene in Bulgaria. The AES-92 is certified as meeting EUR, and its V-392 reactor is considered state of the art. They have four coolant loops and are rated 3000 MWt.
AES-2006, MIR-1200, VVER-TOI
A third-generation standardised VVER-1200 (V-392M and V-491) reactor of 1200 MWe gross and 3200 MWt is in the AES-2006 plant. Novovoronezh units are expected to provide 1068 MWe net, 1170 MWe net is quoted elsewhere, possibly for Leningrad. It is an evolutionary development of the well-proven VVER-1000 in the AES-92 plant, with longer life (60 years for non-replaceable equipment), greater power, and greater efficiency (34.8% net instead of 31.6%) and up to 70 GWd/t burn-up. It retains four coolant loops.
The lead units are being built at Novovoronezh II (V-392M) and Leningrad II (V-491), to start operation in 2014. An AES-2006 plant will consist of two of these OKB Gidropress reactor units expected to run for 50-60 years with capacity factor of 90%. Overnight capital cost was said to be US$ 1200/kW (though the first contract was about $2100/kW) and serial 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-catcher.
While Gidropress is responsible for the actual 1200 MWe reactor, Moscow AEP and St Petersburg AEP are going different ways on the cooling systems, and one or the other may be chosen for future plants once Leningrad II and Novovoronezh II are operating. Passive safety systems prevail in Moscow's V-392M design, while St Petersburg's V-491 design focuses on active safety systems based on Tianwan V428 design. (Details in Russia NP paper)
Atomenergoproekt say that the AES-2006 conforms to both Russian standards and European Utilities Requirements (EUR). In Europe the basic technology is being called the Europe-tailored reactor design, MIR-1200 (Modernised International Reactor) with some Czech involvement. Those bid for Temelin are quoted as 1158 MWe gross,1078 MWe net, In 2010 Atomenergoproekt announced the VVER-TOI(typical optimised, with enhanced information) design with upgraded pressure vessel, increased power to 3300 MWt and 1255-1300 MWe gross (nominally 1300), improved core design to increase cooling reliability, further development of passive safety with 72-hour grace period requiring no operator intervention after shutdown, lower construction and operating costs, and 40-month construction time. It will use a low-speed turbine-generator. The project was initiated in 2009 and is due to be complete and submitted for EUR accreditation in December 2012. In October 2011 the design was reported as half complete. In June 2012 Rosatom said it would apply for VVER-1200 design certification in UK and USA, through Rusatom Overseas, probably with the VVER-TOI version.
The VVER-1500 model was being developed by Gidropress. It will have enhanced safety, giving 1500 MWe gross
from 4250 MWt. Design was expected to be complete in 2007 but the project was shelved in 2006 in favour of the evolutionary VVER-1200. It remains a 4-loop design, with increased pressure vessel diameter to 5 metres, 241 fuel assemblies in core enriched to 4.4%, burn-up 45-55 and up to 60 GWd/t and life of 60 years. If revived, it will meet EUR criteria. In China, there are two indigenous designs based on a French predecessor but developed with modern features. In October 2011 CNNC announced that its ACP1000 was entering the engineering design stage, based on Fuqing units 5 & 6, with 1100 MWe nominal power and load-following capability. It has 177 fuel assemblies, 18-month refuelling interval, and a 60-year design life. Its three coolant loops deliver 3060 MWt, inside double containment and having active safety systems with some passive elements. Seismic rating is 300 gal.
In parallel, China Guangdong Nuclear Power Corporation (CGNPC) led the development of the ACPR1000, a 3-loop unit of about 1100 MWe with 157 fuel assemblies (same as French predecessor), double containment and core-catcher. Its safety attributes comply with international requirements and it has a design life of 60 years.
Another US-origin but international project which is a few years behind the AP1000 is the IRIS (International Reactor Innovative & Secure). 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 burnable poison, in fact fuel assemblies are "identical to those ... in the AP1000". These would have burn-up of 60 GWd/t with fuelling interval of 3 to 3.5 years, but IRIS is designed ultimately for fuel with 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. Estonia once expressed interest in building a pair of them. Some consortium partners are interested in desalination, one in district heating.
OKBM's VBER-300 PWR is a 295-325 MWe unit (917 MWt) 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.
The Reduced-Moderation Water Reactor (RMWR) is a light water reactor, essentially as used today, with the fuel packed in more tightly to reduce the moderating effect of the water. Considering the BWR variant (resource-renewable BWR - RBWR), only the fuel assemblies and control rods are different. In particular, the fuel assemblies are much shorter, so that they can still be cooled adequately. Ideally they are hexagonal, with Y-shaped control rods. The reduced moderation means that more fissile plutonium is produced and the breeding ratio is around 1 (instead of about 0.6), and much more of the U-238 is converted to Pu-239 and then burned than in a conventional reactor. Burn-up is about 45 GWd/t, with a long cycle. Initial seed (and possibly all) MOX fuel needs to have about 10% Pu. The void reactivity is negative, as in conventional LWR. A Hitachi RBWR design based on the ABWR-II has the central part of each fuel assembly (about 80% of it) with MOX fuel rods and the periphery uranium oxide. In the MOX part, minor actinides are burned as well as recycled plutonium.
The main rationale for RMWRs is extending the world's uranium resource and providing a bridge to widespread use of fast neutron reactors. Recycled plutonium should be used preferentially in RMWRs rather than as MOX in conventional LWRs, and multiple recycling of plutonium is possible. Japan Atomic Energy Research Institute (JAERI) started the research on RMWRs in 1997 and then collaborated in the conceptual design study with the Japan Atomic Power Company (JAPCO) in 1998. Hitachi have also been closely involved.
A new reprocessing technology is part of the RMWR concept. This is the fluoride volatility process, developed in 1980s, and is coupled with solvent extraction for plutonium to give the Fluorex process. In this, 90-92% of the uranium in the used fuel is volatalised as UF6, then purified for enrichment or storage. The residual is put through a Purex circuit which separates fission products and minor actinides as high-level waste, leaving the unseparated U-Pu mix (about 4:1) to be made into MOX fuel.
Heavy Water Reactors
(moderated and mostly cooled by heavy water)
In Canada, the
government-owned Atomic Energy of Canada Ltd (AECL) had two designs under
development which are based on its reliable CANDU-6 reactors, the most recent of
which are operating in China. In 2011 the reactor division of AECL was sold and
became Candu Energy Inc, a subsidiary of SNC-Lavalin.
The CANDU-9 (925-1300 MWe) was developed from the
CANDU-6 also as a single-unit plant. It had flexible fuel requirements which
have been taken forward to the EC6. A two year licensing review of the CANDU-9
design was successfully completed early in 1997, but the design has been
Some of the innovation of the CANDU-9, along with
experience in building recent Korean and Chinese units, was then put back into
the Enhanced CANDU-6 (EC6). This is
to be built as twin units - with power increase to 740-750 MWe gross (690 MWe
net, 2084 MWt) and flexible fuel options, plus 4.5 year construction and 60-year
plant life (with mid-life pressure tube replacement). EC6 is presented as a
third-generation design and is under consideration for new build in Ontario and
overseas. Phase 2 of CNSC’s vendor pre-project design review was completed in
2012, with phase 3 on target for early 2013.
Versatility of fuel is a claimed feature of the EC6
and its derivatives. As well as natural uranium, it can use direct recovered/
reprocessed uranium (RU) from used PWR fuel, natural uranium equivalent (NUE –
DU + RU), MOX (DU + Pu), fertile fuels such as LEU + thorium and Th with Pu, and
closed cycle fuels (Th + U-233 + Pu). The NUE fuel cycle with fill-core NUE is
being demonstrated at Qinshan in China in CANDU-6 units*. There is also a
program for the Advanced Fuel Candu Reactor (AFCR) – an adaptation of EC6 - on
direct use of RU- and also LEU + thorium-based CANDU fuel. Finally a CANMOX fuel
is proposed with EC6 for disposal of the UK’s plutonium stock.
* RU with 0.9% U-235 plus DU
gives 0.7% NUE, which is burned down to about 0.25% U-235.
The Advanced Candu Reactor (ACR), a 3rd generation reactor design, was a more innovative concept, but has now been shelved. 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 (3200 MWt) became the focus of attention by AECL (now Candu Energy Inc). 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-1000 was moving towards design certification in Canada, and a 3-phase vendor pre-project design review was completed in 2010. In 2007 AECL applied for UK generic design assessment (pre-licensing approval) but then withdrew after the first stage. All licensing progress has ceased.
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 gross (284 MWe net, 920 MWt) reactor moderated by heavy water at low pressure. The calandria has about 450 vertical pressure tubes and the coolant is boiling light water circulated by convection. A large heat sink - "Gravity-driven water pool" - with 7000 cubic metres of water is near the top of the reactor building. 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, with two thirds of that energy coming from thorium via U-233.
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.
As well as U-233, some U-232 is formed, and the highly gamma-active daughter products of this confer a substantial proliferation resistance.
In 2009 an export version of this design was announced: the AHWR-LEU. This will use low-enriched uranium plus thorium as a fuel, dispensing with the plutonium input. About 39% of the power will come from thorium (via in situ conversion to U-233), and burn-up will be 64 GWd/t. Uranium enrichment level will be 19.75%, giving 4.21% average fissile content of the U-Th fuel. While designed for closed fuel cycle, this is not required. Plutonium production will be less than in light water reactors, and the fissile proportion will be less and the Pu-238 portion three times as high, giving inherent proliferation resistance. The AEC says that "the reactor is manageable with modest industrial infrastructure within the reach of developing countries."
In the AHWR-LEU, the fuel assemblies will be configured:
Inner ring: 12 pins Th-U with 3.555% U-235,
Intermediate ring: 18 pins Th-U with 4.345% U-235,
Outer ring: 24 pins Th-U with 4.444% U-235.
High-Temperature Gas-Cooled Reactors
These reactors use helium as a coolant at up to 950ºC, which either makes steam conventionally or directly 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 as hexagonal 'prisms' of graphite, or in billiard ball-sized pebbles of graphite encased in silicon carbide.
The first commercial version will be China's HTR-PM, being built at Shidaowan in Shandong province. It has been developed by Tsinghua University's INET, which is the R&D leader and Chinergy Co., with China Huaneng Group leading the demonstration plant project. This will have two reactor modules, each of 250 MWt/ 105 MWe, using 9% enriched fuel (520,000 elements) giving 80 GWd/t discharge burnup. With an outlet temperature of 750ºC the pair will drive a single steam cycle turbine at about 40% thermal efficiency. This 210 MWe Shidaowan demonstration plant is to pave the way for an 18-unit (3x6x210MWe) full-scale power plant on the same site, also using the steam cycle. Plant life is envisaged as 60 years with 85% load factor.
South Africa's Pebble Bed Modular Reactor (PBMR) was being developed by a consortium led by the utility Eskom, with Mitsubishi Heavy Industries from 2010. It drew on German expertise and aimed for a step change in safety, economics and proliferation resistance. Production units would be 165 MWe. The PBMR would ultimately have a direct-cycle (Brayton cycle) gas turbine generator and thermal efficiency about 41%, the helium coolant leaving the bottom of the core at about 900°C and driving a turbine. Power is adjusted by changing the pressure in the system. The helium is passed through a water-cooled pre-cooler and intercooler before being returned to the reactor vessel. (In the Demonstration Plant it would transfer heat in a steam generator rather than driving a turbine directly.) However, development has ceased due to lack of funds and customers.
A larger US design, the Gas Turbine - Modular Helium Reactor (GT-MHR), is planned 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 OKBM Afrikantov, supported by Fuji (Japan). Initially it was to be used to burn pure ex-weapons plutonium at Seversk (Tomsk) in Russia. The preliminary design stage was completed in 2001, but the program has stalled since. In February 2010 General Atomics announced its Energy Multiplier Module (EM2) design, based on the GT-MHR.
Areva's Antares is based on the GT-MHR.
Fuller descriptions of HTRs is in the Small Nuclear Power Reactors paper .
Fast Neutron Reactors
(not moderated, cooled by liquid metal)
Several countries have research and development programs for improved Fast Breeder Reactors (FBR), which are Fast Neutron Reactors (FNR) configured with a conversion or breeding ratio of more than 1 (ie more fissile nuclei are produced than are fissioned). 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 have supplied electricity commercially. About 300 reactor-years of operating experience have been accumulated. See also Fast Neutron Reactors paper.
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 almost ceased for some years, and that on the 1450 MWe European FBR has apparently lapsed. 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 2012, fuelled with uranium-plutonium oxide (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 for 15 years due to a sodium leak. It restarted in 2010.
Mitsubishi Heavy Industries (MHI) is involved with a consortium to build the Japan Standard Fast Reactor (JSFR) concept, though with breeding ratio less than 1:1. This is a large unit which will burn actinides with uranium and plutonium in oxide fuel. It could be of any size from 500 to 1500 MWe. In this connection MHI has also set up Mitsubishi FBR Systems (MFBR). The demonstration FR model is due to be committed in 2015 and on line in 2025, and a 1500 MWe commercial FR is proposed by MHI for 2050.
The Russian BN-600 fast breeder reactor at Beloyarsk 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.
Advanced FNRs include the following:
The first BN-800, a new more powerful (880 MWe gross, 2100 MWt) FBR from OKBM with improved features is being built at Beloyarsk. It has considerable fuel flexibility - U+Pu nitride, MOX, or metal, and with breeding ratio up to 1.3. With three loops, outlet primary coolant temperature is 547ºC and steam temperature 470ºC. Service life is 40 years. It has much enhanced safety and improved economy - while capital cost is 20% more than VVER-1200, operating cost is expected to be only 15% more than VVER. It is capable of burning 1.7 to 2 tonnes of plutonium per year from dismantled weapons and will test the recycling of minor actinides in the fuel. The BN-800 has been sold to China, and two units are due to start construction there in 2012.
However, the BN-800 is likely to be the last such reactor design built (outside India’s thorium program), with a fertile blanket of depleted uranium around the core. Further fast reactors will have an integrated core to minimise the potential for weapons proliferation from bred Pu-239.
The BN-1200 is being designed by OKBM for operation with MOX fuel from 2020 and dense nitride U-Pu fuel subsequently, in closed fuel cycle. Rosatom plans to submit the BN-1200 to the Generation IV International Forum (GIF) as a Generation IV design. The BN-1200 will produce 2900 MWt (1220 MWe), has a 60-year design life, simplified refuelling, and burn-up of up to 120 GWd/t. Intermediate heat exchanger temperature 550°C, with 527°C in secondary sodium circuit and 510°C outlet at 14 MPa from steam generators. Lead coolant is also a possibility. Thermal efficiency is 42% gross, 39% net. It is expected to have a breeding ratio of 1.2 initially and up to 1.35 for MOX fuel, and then 1.45 for nitride fuel. Fuel burn-up is designed to progress from 14.3% to 21%. It will have 426 fuel assemblies and 174 radial blanket assemblies surrounded by 599 boron shielding assemblies. The capital cost is expected to be much the same as VVER-1200. OKBM envisages about 11 GWe of such plants by 2030, possibly including South Urals NPP. Design is expected to be complete in 2014, and tentative plans are for construction of the first unit at Beloyarsk from 2015 with commercial operation from 2020. It is intended to produce electricity at RUR 0.65/kWh (US 2.23 cents/kWh). This is part of a federal Rosatom program, the Proryv (Breakthrough) Project.
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 from NIKIET is the BREST fast neutron reactor, of 300 MWe (700 MWt) or more with lead as the primary coolant, at 540ºC, 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). Also it is an equilibrium core, so there are no spare neutrons to irradiate targets. 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 by 2020, and 1200 MWe (2800 MWt) units are proposed. Both designs have two cooling loops.
The European Lead-cooled SYstem (ELSY) of 600 MWe in Europe, led by Ansaldo Nucleare from Italy and financed by Euratom. ELSY is a flexible fast neutron reactor which can use depleted uranium or thorium fuel matrices, and burn actinides from LWR fuel. Liquid metal (Pb or Pb-Bi eutectic) cooling is at low pressure .The design was nearly complete in 2008 and a small-scale demonstration facility is planned. It runs on MOX fuel at 480°C and the molten lead is pumped to eight steam generators, though decay heat removal is passive, by convection.
In the USA, GE was involved in designing a modular liquid metal-cooled inherently-safe reactor - PRISM. GE with the DOE national laboratories were developing PRISM during the advanced liquid-metal fast breeder reactor (ALMR) program. No US fast neutron reactor has so far been larger than 66 MWe and none has supplied electricity commercially.
Today's PRISM is a GE-Hitachi design for compact modular pool-type reactors with passive cooling for decay heat removal. After 30 years of development it represents GEH's Generation IV solution to closing the fuel cycle in the USA. Each PRISM Power Block consists of two modules of 311 MWe each, operating at high temperature - over 500°C. The pool-type modules below ground level contain the complete primary system with sodium coolant. The Pu & DU fuel is metal, and obtained from used light water reactor fuel. However, all transuranic elements are removed together in the electrometallurgical reprocessing so that fresh fuel has minor actinides with the plutonium. Fuel stays in the reactor about six years, with one third removed every two years. Used PRISM fuel is recycled after removal of fission products. The commercial-scale plant concept, part of a Advanced Recycling Centre, uses three power blocks (six reactor modules) to provide 1866 MWe. See also electrometallurgical section in Processing Used Nuclear Fuel paper.
A variant of this is proposed to utilise the UK's reactor-grade plutonium stockpile. A pair of PRISM units built at Sellafield would be operated initially so as to bring the material up to the highly-radioactive 'spent fuel standard' of self-protection and proliferation resistance. The whole stockpile could be irradiated thus in five years, with some by-product electricity and the plant would then proceed to re-use that stored fuel over perhaps 55 years solely for 600 MWe of electricity generation.
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.
See also paper on Fast Neutron Reactors.
Generation IV Designs
See paper on six Generation IV Reactors, also DOE paper.
See also paper on Small Nuclear Power Reactors for other advanced designs, mostly under 300 MWe. This paper includes some designs which have become significantly larger than 300 MWe since first being described, but which are outside the mainstream categories dealt with here.
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.
Nuclear Engineering International, various, and 2002 Reactor Design supplement. March 2012: Atmea1 reactor.
ABB Atom Dec 1999; Nukem market report July 2000;
The New Nuclear Power, 21st Century, Spring 2001,
Lauret, P. et al, 2001, The Nuclear Engineer 42, 5.
Smirnov V.S. et al, 2001, Design features of BREST reactors, KAIF/KNS conf.Proc.
OECD NEA 2001, Trends in the Nuclear Fuel Cycle;
Carroll D & Boardman C, 2002, The Super-PRISM Reactor System, The Nuclear Engineer 43,6;
Twilley R C 2002, Framatome ANP's SWR1000 reactor design, Nuclear News, Sept 2002.
Torgerson D F 2002, The ACR-700, Nuclear News Oct 2002.
IEA-NEA-IAEA 2002, Innovative Nuclear Reactor Development
Perera, J, 2003, Developing a passive heavy water reactor, Nuclear Engineering International, March.
Sinha R.K.& Kakodkar A. 2003, Advanced Heavy Water Reactor, INS News vol 16, 1.
US Dept of Energy, EIA 2003, New Reactor Designs.
Matzie R.A. 2003, PBMR - the first Generation IV reactor to be constructed, WNA Symposium.
LaBar M. 2003, Status of the GT-MHR for electricity production, WNA Symposium.
Carelli M 2003, IRIS: a global approach to nuclear power renaissance, Nuclear News Sept 2003.
Perera J. 2004, Fuelling Innovation, IAEA Bulletin 46/1.
AECL Candu-6 & ACR publicity, late 2005.
IAEA Status report 83 – APR1400
Atomenergoproekt web site
Appendix: US Nuclear Regulatory Commission draft policy, May 2008.
The Commission believes designers should consider several reactor characteristics, including:
- Highly reliable, less complex safe shutdown systems, particularly ones with inherent or passive safety features;
- Simplified safety systems that allow more straightforward engineering analysis, operate with fewer operator actions and increase operator comprehension of reactor conditions;
- Concurrent resolution of safety and security requirements, resulting in an overall security system that requires fewer human actions;
- Features that prevent a simultaneous breach of containment and loss of core cooling from an aircraft impact, or that inherently delay any radiological release, and;
- Features that maintain spent fuel pool integrity following an aircraft impact.
Advanced Thermal Reactors being marketed
|Country and developer
||Size MWe gross
(improved safety in all)
||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.
|AP600: NRC certified 1999, FOAKE.
AP1000 NRC certification 2005, under construction in China, many more planned there. Four due to start construction in USA in 2012.
- Simplified construction and operation.
- 3 years to build.
- 60-year plant life.
||Future French standard.
French design approval.
Being built in Finland, France & China. Undergoing certification in USA.
- Evolutionary design.
- High fuel efficiency.
- Flexible operation
||Developed from ABWR,
undergoing certification in USA, likely constructiion there.
- Evolutionary design.
- Short construction time.
|Basic design in progress,
planned for Tsuruga
US design certification application.
- Hybrid safety features.
- Simplified Construction and operation.
(KHNP, derived from Westinghouse)
||Under construction - Shin Kori 3 & 4. Sold to UAE.
- Evolutionary design.
- Increased reliability.
- Simplified construction and operation.
||French design approval Feb 2012, ready for deployment.
- Innovative design.
- High fuel efficiency.
||Under construction at Leningrad, Novovoronezh and Baltic plants
- Evolutionary design.
- High fuel efficiency.
- 50-year plant life
|Canada (Candu Energy)
Licensing approval 1997
- Evolutionary design.
- Flexible fuel requirements.
|China (INET, Chinergy)
||Demonstration plant being built at Shidaowan
- Modular plant, low cost.
- High temperature.
- High fuel efficiency.