# Advanced nuclear power reactors

 Topics:

This EOE article is adapted from an information paper published by the World Nuclear Association (WNA).  WNA information papers are frequently updated, so for greater detail or more up to date numbers, please see the latest version on WNA website (link at end of article).

## Introduction

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 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 reactor designs are still on the drawing board and will not be operational before 2020 at the earliest.

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 These and other second-generation nuclear power units have been found to be safe and reliable, but they are being superseded by better designs.

Reactor suppliers in North America, Japan, Europe, Russia and 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.

Third-generation reactors have:

• a standardized design for each type to expedite licensing, reduce capital cost and reduce construction time;
• a simpler and more rugged design, making them easier to operate and less vulnerable to operational upsets;
• higher availability and longer operating life—typically 60 years;
• reduced possibility of core melt accidents;
• resistance to serious damage that would allow radiological release from an aircraft impact;
• higher burn-up to reduce fuel use and the amount of waste; and
• burnable absorbers ("poisons") to extend fuel life.

The greatest departure from second-generation designs is that many third-generation reactors incorporate passive or inherent safety features that require no active controls or operational intervention to avoid accidents in the event of malfunction, and may rely on gravity, natural convection or resistance to high temperatures. Traditional nuclear reactor safety systems are 'active' in the sense that they involve electrical or mechanical operation on command. Some engineered systems operate passively, e.g., pressure relief valves. 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.

Another departure is that some 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.

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 by about 2010.  These include: ACR from Atomic Energy of Canada Ltd (AECL), IRIS from Westinghouse, PBMR from Eskom and 4S from Toshiba as well as General Atomics' GT-MHR apparently.  See also NRC web page, and Appendix.

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 nuclear reactor and nuclear fuel cycle technology, mostly looking further ahead than the main subjects of this paper:

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

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: GE Hitachi Nuclear Energy (GEH),
- 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 (LWR)

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

One is an advanced boiling water reactor (ABWR) 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 pressurized water reactor (PWR), which was ready for commercialization but is not now being promoted for sale. Eight System 80 reactors in South Korea incorporate many design features of the System 80+, which is the basis of the Korean Next Generation Reactor program, specifically the APR-1400, which is expected to be in operation soon after 2010 and marketed worldwide.

The US Nuclear Regulatory Commission (NRC) gave final design certification for both in May 1997, noting that they exceeded NRC "safety goals by several orders of magnitude". The ABWR has also been certified as meeting European requirements for advanced reactors.

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

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

Separate from the NRC process and beyond its immediate requirements, the US nuclear industry selected one standardized design in each category—the large ABWR and the medium-sized AP-600, for detailed first-of-a-kind engineering (FOAKE) work. The US$200 million program, half funded by DOE, is now complete. It means that prospective buyers now have fuller information on construction costs and schedules. The 1100 MWe Westinghouse AP-1000, scaled-up from the AP-600, received final design approval 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 were originally projected at $1200 per kilowatt and modular design is expected to reduce construction time to 36 months. The AP-1000 generating costs are expected to be very competitive and it has a 60 year operating life. It has been selected for building in China (4 units, 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. A contrast between the 1188 MWe Westinghouse reactor at Sizewell B in the UK and the Generation III+ AP1000 of similar-power illustrates the evolution from Generation II 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. GE Hitachi Nuclear Energy’s ESBWR is a Generation III+ technology that utilizes passive safety features and natural circulation principles and is essentially an evolution from its predecessor design, the SBWR at 670 MWe. The ESBWR will produce approximately 1520 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 and is on schedule for full design certification in 2010-11. It is favoured for early US construction and could be operational in 2015. 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. Another US-origin but international project, a few years behind the AP-1000, is the International Reactor Innovative & Secure (IRIS) project. Westinghouse is leading a wide consortium developing it as an advanced 3rd Generation project. IRIS is a modular 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 review stage with NRC. Estonia has expressed interest in building a pair of them. 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$7 cents/kWh. Two more started up in 2004 & 2005. Several of t 1350 MWe units are under construction in Japan and Taiwan.

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

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 $1500/kW. US design certification application was in January 2008 with approval expected in 2011 and certification mid 2012. The first units may be built for TXU at Comanche Peak near Dallas, Texas. In March 2008 MHI submitted the same design for EUR certification, as EU-APWR. The Atmea joint venture has been established by Areva NP and Mitsubishi Heavy Industries to develop an 1100 MWe (net) three-loop PWR with extended fuel cycles, 37% thermal efficiency and the capacity to use mixed-oxide fuel only. Fuel cycle is 12-24 months and the reactor has load-following capability. They expect to have this ready for licence application by 2010. The reactor is regarded as mid-sized relative to other generation III units and will be marketed primarily to countries embarking upon nuclear power programs. In South Korea, the APR-1400 Advanced PWR design has evolved from the US System 80+ with enhanced safety and seismic robustness and was earlier known as the Korean Next-Generation Reactor. Design certification by the Korean Institute of Nuclear Safety was awarded in May 2003. The first of these 1450 MWe reactors will be Shin-Kori-3 & 4, expected to be operating about 2012. Fuel has burnable poison and will have up to 60 GWd/t burn-up. Projected cost is US$ 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 pressurized water reactor (EPR), which was confirmed in mid-1995 as the new standard design for France and received French design approval in 2004. It is derived from the French N4 and German Konvoi types and is expected to provide power about 10% cheaper than the N4. It will operate flexibly to follow loads, have fuel burn-up of 65 GWd/t and the highest thermal efficiency of any light water reactor, at 36%. 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, the third European one will be at Penly in France, and two further units are planned in China. A US version, the US-EPR, was submitted for US design certification in December 2007, and this is expected to be granted early 2012. 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 1000-1290 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 nuclear reactor design 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 the AES-2006 plant is an evolutionary development of the well-proven VVER-1000 in the AES-92 plant, with longer life (50, not 30 years), greater power, and greater efficiency (36.56% instead of 31.6%).  The lead units are being 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%.  Overnight capital cost was 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

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

The CANDU-9 (925-1300 MWe) was also developed from the CANDU-6 as a single-unit plant. It has flexible fuel requirements ranging from natural uranium through slightly-enriched uranium, recovered uranium from reprocessing spent pressurized water reactor (PWR) fuel, mixed oxide (MOX or U & Pu) fuel, direct use of spent PWR fuel, to thorium. It may be able to burn military plutonium or actinides separated from reprocessed PWR/BWR waste. A two-year licensing review of the CANDU-9 design was successfully completed in early 1997, 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 third-generation reactor, is more of a innovative concept. While retaining the low-pressure heavy water moderator, it incorporates some features of the pressurized water reactor. Adopting light water cooling and a more compact core reduces capital cost, and because the reactor is run at higher temperature and coolant pressure, it has higher thermal efficiency.

The ACR-700 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) but then withdrew after the first stage.  In the USA, the ACR-700 is listed by NRC as being at pre application review stage.  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 (e.g., 25 MPa and 625°C) to provide 40% thermal efficiency. The size range envisaged is 350 to 1150 MWe, depending on the number of fuel channels used. Commercialization envisaged after 2020.

India is developing the Advanced Heavy Water reactor (AHWR) as the third stage in its plan to utilize thorium to fuel its overall nuclear power program. The AHWR is a 300 MWe nuclear reactor moderated by heavy water at low pressure. The calandria has 500 vertical pressure tubes and the coolant is boiling light water circulated by convection. Each fuel assembly has 30 thorium-uranium-233 oxide pins and 24 plutonium-thorium oxide pins around a central rod with burnable absorber. Burn-up of 24 GWd/t is envisaged. It is designed to be self-sustaining in relation to uranium-233 bred from thorium-232 and have a low plutonium inventory and consumption, with slightly negative void coefficient of reactivity. It is designed for 100 year plant life and is expected to utilise 65% of the energy of the fuel.

## High-Temperature Gas-Cooled Reactors

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

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

South Africa's Pebble Bed Modular Reactor (PBMR) is being developed by a consortium led by the utility Eskom, and drawing on German expertise.It aims for 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 center column.

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

Overnight capital cost (when in clusters of eight units) is expected to be modest and generating cost very competitive.  The PBMR project has reverted to Eskom and is funded by the South African government.  A demonstration plant is due to be built in 2009, with fuel loading expected in 2013. In the USA, PBMR Ltd is planning to submit a design certification application for the reactor in 2008, and to bid for a nuclear-powered thermochemical hydrogen production plant based on it at the Idaho National Laboratory.

A larger US design, the Gas Turbine-Modular Helium Reactor (GT-MHR), will be built as modules of 285 MWe each directly driving a gas turbine at 48% thermal efficiency. The cylindrical core consists of 102 hexagonal fuel element columns of graphite blocks with channels for helium and control rods. Graphite reflector blocks are both inside and around the core. Half of the core is replaced every 18 months. Burn-up is about 100,000 MWd/t. The GT-MHR is being developed by General Atomics in partnership with Russia's Minatom, supported by Fuji (Japan). Initially, it will be used to burn pure ex-weapons plutonium at Tomsk in Russia. The preliminary design stage was completed in 2001.

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

## Fast Neutron Reactors

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

About 20 liquid metal-cooled FBRs have already been operating, some since the 1950s, and some supply electricity commercially. About 300 reactor-years of operating experience have been accumulated.

Natural uranium contains about 0.7% 235U and 99.3 % 238U. In any reactor, the 238U component is turned into several isotopes of plutonium during its operation. Two of these, plutonium-239 (239Pu) and 241Pu, then undergo fission in the same way as 235U to produce heat. In a fast neutron reactor, this process is optimized so that it can 'breed' fuel, often using a depleted uranium blanket around the core. FBRs can utilize uranium at least 60 times more efficiently than a normal reactor. They are, however, expensive to build and could only be justified economically if uranium prices are high.

For this reason, research work on the 1450 MWe European FBR has almost ceased. Closure of the 1250 MWe French Superphenix FBR after very little operation over 13 years also set back developments.

Research continues in India.  At the Indira Gandhi Centre for Atomic Research, a 40 MWt fast breeder test reactor has been operating since 1985. In addition, the tiny Kamini reactor at the Centre is employed to explore the use of thorium as nuclear fuel by breeding fissile uranium-233 (233U). In 2004, construction of a 500 MWe prototype fast breeder reactor started at Kalpakkam. The unit is expected to be operating in 2011, fuelled with uranium-plutonium carbide (the reactor-grade Pu being from its existing PHWRs) and with a thorium blanket to breed fissile 233U. This will take India's ambitious thorium program to stage 2, and set the scene for eventual full utilization of the country's abundant thorium resources as fuel for nuclear reactors.

Japan plans to develop FBRs, and its Joyo experimental reactor which has been operating since 1977 is now being boosted to 140 MWt. The 280 MWe Monju prototype commercial FBR was connected to the grid in 1995, but was subsequently shut down due to a sodium leak. It is planned to restart it in 2009.

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

Construction has started at Beloyarsk on the first BN-800, a new larger (880 MWe) FBR from OKBM with improved features including fuel flexibility—U+Pu nitride, MOX, or metal—and with breeding ratio up to 1.3. It has much enhanced safety and improved economy with operating costs expected to be only 15% more relative to VVER. It is capable of burning 2 tonnes of plutonium per year from dismantled weapons and will test the recycling of minor actinides in the fuel.

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

In the USA, GE was involved in designing a modular 150 MWe liquid metal-cooled inherently-safe reactor—PRISM. GE 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.  Modules are 200 to 360 MWe and operate at high temperature – over 500°C.  The pool-type modules 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.  The commercial-scale plant concept uses six reactor modules to provide 1200 to 2200 MWe.

Korea's KALIMER (Korea Advanced LIquid MEtal Reactor) is a 600 MWe pool type sodium-cooled fast reactor designed to operate at over 500°C.  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.

## Accelerator-Driven Systems

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

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

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

## 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.

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Glossary

### Citation

Hore-Lacy, I., & Association, W. (2010). Advanced nuclear power reactors. Retrieved from http://www.eoearth.org/view/article/149847