# Small nuclear power reactors

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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

As nuclear power generation has become established since the 1950s, the size of reactor units has grown from 60 MWe to more than 1600 MWe, with corresponding economies of scale in operation. At the same time there have been many hundreds of smaller reactors built both for naval use (up to 190 MW thermal) and as neutron sources, yielding enormous expertise in the engineering of deliberately small units.

Today, due partly to the high capital cost of large nuclear power reactors generating electricity via the steam cycle and partly to the need to service small electricity grids under about 4 GWe, there is a move to develop smaller units. (A very general rule is that no single unit should be larger than 15% of grid capacity.)  These smaller units may be built independently or as modules in a larger complex, with capacity added incrementally as required. Economies of scale are provided by the numbers produced. There are also moves to develop small units for remote sites. The IAEA defines "small" as under 300 MWe, but in general today 500 MWe might be considered an upper limit to "small".  The IAEA is reported to project up to 1000 small nuclear reactors producing power by 2040.

The most prominent modular project is the South African-led consortium developing the Pebble Bed Modular Reactor (PBMR) of 170 MWe. In China, Chinergy is preparing to build a similar unit, the 195 MWe HTR-PM. A US-led group is developing another design with 285 MWe modules. Both US types drive gas turbines directly, using helium as a coolant and operating at very high temperatures. They build on the experience of several innovative reactors developed in the 1960s and 1970s.

Another significant line of development is in very small fast reactors of under 50 MWe.

Generally, modern small reactors for power generation are expected to have greater simplicity of design, economy of mass production, and reduced siting costs. Many are also designed for a high level of passive or inherent safety in the event of malfunction. Traditional reactor safety systems are 'active' in the sense that they involve electrical or mechanical operation on command. Some engineered systems operate passively, e.g., pressure relief valves. Both require parallel redundant systems. Inherent or full 'passive' safety depends only on physical phenomena such as convection, gravity or resistance to high temperatures, not on functioning of engineered components.

Some small reactors are conceived for areas away from transmission grids and with small loads, others are designed to operate in clusters in competition with large units. The cost of electricity from a 50 MWe unit is estimated by the U.S. Department of Energy (DOE) as 5.4 to 10.7 cents/kWh (compared with charges in Alaska and Hawaii from 5.9 to 36.0 c/kWh).

US Congress is now funding research on both small modular nuclear power plants (assembled on site from factory-produced modules) and advanced gas-cooled designs (which are modular in the sense that up to ten or more units are progressively built to comprise a major power station).

Already operating in a remote corner of Siberia are four small units at the Bilibino co-generation plant. These four 62 MWt (thermal) units are an unusual graphite-moderated boiling water reactor (BWR) design with water/steam channels through the moderator. They produce steam for district heating and 11 MWe (net) electricity each. They have performed well since 1976, much more cheaply than fossil fuel alternatives in the Arctic region.

## Light Water Reactors (LWRs)

US experience with small light water reactors (LWRs) has been of very small military power plants, such as the 11 MWt, 1.5 MWe (net) PM-3A nuclear reactor which operated at McMurdo Sound in Antarctica 1962-72, generating a total of 78 million kWh. There was also an Army program for small reactor development and some successful small reactors from the main national program commenced in the 1950s. One was the Big Rock Point BWR of 67 MWe, which operated for 35 years to 1997.

Of the following, the KLT and VBER designs have conventional pressure vessel plus external steam generators (PV/loop design). The others mostly have the steam supply system inside the reactor pressure vessel ('integral' PWR design). All have enhanced safety features relative to current PWRs.

The Russian KLT-40S is a reactor well-proven in icebreakers and now proposed for wider use in desalination and, on barges, for remote area power supply. Here a 150 MWt unit produces 35 MWe (gross) as well as up to 35 MW of heat for desalination or district heating (or 38.5 MWe gross if power only). These are designed to run 3-4 years between refuelling and it is envisaged that they will be operated in pairs to allow for outages (70% capacity factor), with on-board refuelling capability and spent fuel storage. At the end of a 12-year operating cycle the whole plant is taken to a central facility for overhaul and storage of spent fuel. Two units will be mounted on a 20,000 tonne barge.

Although the reactor core is normally cooled by forced circulation, the OKBM design relies on convection for emergency cooling. Fuel is uranium aluminium silicide with enrichment levels of up to 20%, giving up to 4-year refuelling intervals.

OKBM is developing a new icebreaker reactor – RITM-200 – to replace the KLT reactors and to serve in floating nuclear power plants.  This is an integral 210 MWt, 55 MWe PWR with inherent safety features.  A single compact RITM-200 could replace twin KLT-40S (but yielding less total power).  A major challenge is the reliability of steam generators and associated equipment which is much less accessible when inside the reactor pressure vessel.

A larger Russian factory-built and barge-mounted unit (requiring a 12,000 tonne vessel) is the VBER-150, of 350 MW thermal, 110 MWe. It has modular construction and is derived by OKBM from naval designs, with two steam generators. Uranium oxide fuel enriched to 4.7% has burnable poison; it has low burnup (31 GWd/t average, 41.6 GWd/t max) and 8 year refuelling interval.

OKBM's larger VBER-300 pressurized water reactor (PWR) is a 295 MWe unit, the first of which will be built in Kazakhstan.  It was originally envisaged in pairs as a floating nuclear power plant, displacing 49,000 tonnes. As a cogeneration plant, it is rated at 200 MWe and 1900 GJ/hr. The reactor is designed for 60 year life and 90% capacity factor. It has four steam generators and a cassette core with 85 fuel assemblies enriched to 5% and 48 GWd/tU burn-up. Versions with three and two steam generators are also envisaged, of 230 and 150 MWe respectively. Also with more sophisticated and higher-enriched (18%) fuel in the core, the refuelling interval can be pushed from 2 years out to 15 years with burn-up to 125 GWd/tU.  A 2006 joint venture between Atomstroyexport and Kazatomprom sets this up for development as a basic power source in Kazakhstan, then for export.

Another larger Russian reactor is the VK-300 boiling water reactor (BWR) being developed specifically for cogeneration of both power (250 MWe) and desalination (150 MWe plus 1675 GJ/hr) by the Research & Development Institute of Power Engineering (NIKIET).  It has evolved from the VK-50 BWR at Dimitrovgrad, but uses standard components wherever possible, e.g., the reactor vessel of the VVER-1000. Fuel burn-up is 41 GWday/tU. It is capable of producing 250 MWe if solely electrical.  In September 2007 it was announced that six would be built at Kola and at Primorskaya in the far east, to start operating 2017-20.

A smaller Russian OKBM PWR unit under development is the ABV, with a range of sizes from 45 MW thermal (ABV-6M ) down to 18 MWt (ABV-3), giving 4-18 MWe outputs.  The units are compact, with integral steam generator.  The whole unit will be factory-produced for ground or barge mounting – the ABV-6M would require a 3500 tonne barge, the ABV-3: 1600 tonne.  The core is similar to that of the KLT-40 except that enrichment is 16.5% and average burnup 95 GWd/t.  Refuelling interval is about 8-10 years, and service life about 50 years.

The CAREM (advanced small nuclear power plant) being developed by CNEA and INVAP in Argentina is a modular 100 MWt/ 27 MWe pressurized water reactor (PWR) with integral steam generators designed to be used for electricity generation (27 MWe or up to 100 MWe) or as a research reactor or for water desalination (with 8 MWe in cogeneration configuration). CAREM has its entire primary coolant system within the reactor pressure vessel, self-pressurized and relying entirely on convection. Fuel is standard 3.4% enriched PWR fuel, with burnable poison, and is refuelled annually. It is a mature design that could be deployed within a decade.

On a larger scale, South Korea's SMART (System-integrated Modular Advanced Reactor) is a 330 MWt pressurized water reactor with integral steam generators and advanced safety features. It is designed for generating electricity (up to 100 MWe) and/or thermal applications such as seawater desalination. The design life is 60 years, with a 3-year refuelling cycle.

Small-medium reactors with development well advanced
CAREM 27 MWe PWR CNEA & INVAP, Argentina
KLT-40 35 MWe PWR OKBM, Russia
MRX 30-100 MWe PWR JAERI, Japan
IRIS-100 100 MWe PWR Westinghouse-led, international
SMART 100 MWe PWR KAERI, S. Korea
NP-300 100-300 MWe PWR Technicatome (Areva), France
VK-300 300 MWe BWR Atomenergoproekt, Russia
PBMR 165 MWe HTGR Eskom, South Africa, et al
GT-MHR 285 MWe HTGR General Atomics (USA), Minatom (Russia) et al
BREST 300 MWe LMR RDIPE (Russia)
FUJI 100 MWe MSR ITHMSO, Japan-Russia-USA

The Japan Atomic Energy Research Institute (JAERI) is developing the MRX, a small (50-300 MWt) integral PWR reactor for marine propulsion or local energy supply (30 MWe). The entire plant would be factory-built. It has conventional 4.3% enriched PWR uranium oxide fuel with a 3.5-year refuelling interval and has a water-filled containment to enhance safety. It could be deployed within a decade.

Technicatome (Areva) in France has developed the NP-300 PWR from submarine power plants and aimed it at export markets for power, heat and desalination. It can be built for applications of 100 to 300 MWe or more with up to 500,000 m3/day desalination. The nuclear part is below ground level.

The Chinese NHR-200 is a simple and robust 200 MWt integral PWR design for district heating or desalination. It runs at lower temperature than the above designs. Spent fuel is stored around the core in the pressure vessel.

The International Reactor Innovative & Secure (IRIS) is being developed by Westinghouse as an advanced 3rd-generation reactor. A small version of IRIS is a modular 100 MWe or more pressurized water reactor with integral primary coolant system and circulation by convection. Fuel is similar to present LWRs. Enrichment is 5% with burnable poison and fuelling interval possibly of 5 years (or longer with higher enrichment). IRIS could be built by 2015 if developments proceed. See article on advanced nuclear power reactors for information on larger IRIS.

A similar unit is the NuScale multi-application small PWR which is apparently similar to IRIS but with natural circulation.  It is about 150 MW thermal or 45 MWe, factory-built and with 3 metre diameter pressure vessel and convection cooling.  The whole unit is installed below grade, and eight modules together might form a 360 MWe power station.  An application for US design certification is expected in 2010, with pre-application meetings from mid 2008.  The company was spun out of Oregon Sate University in 2007.

The TRIGA Power System is a pressurized water reactor (PWR) concept based on General Atomics' well-proven research reactor design. It is conceived as a 64 MWt, 16.4 MWe pool-type system operating at a relatively low temperature. The secondary coolant is organic perfluorocarbon. The fuel is uranium-zirconium hydride enriched to 20% and with a little burnable poison and requiring refuelling every 18 months. Spent fuel is stored inside the reactor vessel.

## High-temperature Gas-cooled Reactors (HTR)

Building on the experience of several innovative reactors built in the 1960s and 1970s, new high-temperature gas-cooled reactors (HTGRs) are being developed that will be capable of delivering high-temperature (up to 950°C) helium, either for industrial application via heat exchanger or directly to drive gas turbines for electricity (the Brayton cycle) with almost 50% thermal efficiency possible (efficiency increases 1.5% with each 50°C increment). Technology developed in the last decade makes HTRs more practical than in the past, though the direct cycle means that there must be high integrity of fuel and reactor components.

Fuel for these nuclear reactors is in the form of TRISO particles less than a millimeter in diameter. Each has a kernel (~ 0.5 mm) of uranium oxycarbide, with the uranium enriched up to 20% uranium-235 (235U), though normally less. This is surrounded by layers of carbon and silicon carbide, giving a containment for fission products which is stable to 1600°C or more.  With negative temperature coefficient of reactivity (the fission reaction slows as temperature increases) and passive decay heat removal, this makes the reactors inherently safe.  They do not require any containment building for safety.

The reactors are sufficiently small to allow factory fabrication, and will usually be installed below ground level.

There are two ways in which these TRISO fuel particles are arranged: in blocks—hexagonal 'prisms' of graphite—or in billiard ball-sized pebbles of graphite encased in silicon carbide, each with about 15,000 fuel particles and 9g uranium.  There is a greater amount of spent fuel than from the same capacity in a light water reactor (LWR).  The moderator is graphite.

The Japan Atomic Energy Research Institute's (JAERI) High-Temperature Test Reactor (HTTR) of 30 MW thermal started up at the end of 1998 and has been run successfully at 850°C. In 2004 it achieved 950°C outlet temperature. Its fuel is in 'prisms' and its main purpose is to develop thermochemical means of producing hydrogen from water.

Based on the HTTR, JAERI is developing the Gas Turbine High Temperature Reactor (GTHTR) of up to 600 MW thermal per module. It uses improved HTTR fuel elements with 14% enriched uranium achieving high burn-up (112 GWd/t). Helium at 850°C drives a horizontal turbine at 47% efficiency to produce up to 300 MWe. The core consists of 90 hexagonal fuel columns 8-meters-high arranged in a ring, with reflectors. Each column consists of eight one-meter-high elements 0.4 m across and holding 57 fuel pins made up of fuel particles with 0.55 mm diameter kernels and 0.14 mm buffer layer. In each 2-yearly refuelling, alternate layers of elements are replaced so that each remains for 4 years.

China's HTR-10, a small high-temperature pebble-bed gas-cooled experimental reactor at the Institute of Nuclear and New Energy Technology (INET) at Tsinghua University north of Beijing started up in 2000 and reached full power in 2003. It has its fuel as a 'pebble bed' (27,000 elements) of oxide fuel with average burnup of 80 GWday/tU. Each pebble fuel element has 5g of uranium enriched to 17% in around 8300 particles. The reactor operates at 700°C (potentially 900°C) and has broad research purposes. Eventually it will be coupled to a gas turbine, but meanwhile it has been driving a steam turbine.

Construction of a larger version, the 200 MWe HTR-PM, was approved in principle in November 2005, with construction starting in 2009. This will have two reactors modules, each of 250 MWt, 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.  The size was reduced to 250 MWt from earlier 458 MWt modules in order to retain the same core configuration as the prototype HTR-10 and avoid moving to an annular design like South Africa's PBMR.  This Shidaowan demonstration reactor at Rongcheng in Shandong province is to pave the way for an 18-unit (3x6x200MWe) full-scale power plant on the same site at Weihei, also using the steam cycle.  Plant life is envisaged as 60 years with 85% load factor.

China Huaneng Group, one of China's major generators, is the lead organization involved in the demonstration unit with 47.5% share; China Nuclear Engineering & Construction (CNEC) will have a 32.5% stake and Tsinghua University's INET 20% - it being the main R&D contributor.  Projected cost is US$385 million (but later units falling to US$1500/kW with generating cost about 5c/kWh).  Start-up is scheduled for 2013.  The HTR-PM rationale is both eventually to replace conventional reactor technology for power, and also to provide for future hydrogen production. INET is in charge of R&D, and is aiming to increase the size of the 250 MWt module and also utilise thorium in the fuel.  Eventually a series of HTRs, possibly with Brayton cycle directly driving the gas turbines, will be factory-built and widely installed throughout China.

In 2004, the small HTR-10 reactor was subject to an extreme test of its safety when the helium circulator was deliberately shut off without the reactor being shut down. The temperature increased steadily, but the physics of the fuel meant that the reaction progressively diminished and eventually died away over three hours. At this stage, a balance between decay heat in the core and heat dissipation through the steel reactor wall was achieved and the temperature never exceeded a safe 1600°C. This was one of six safety demonstration tests conducted. The high surface area relative to volume, and the low power density in the core, will also be features of the full-scale units (which are nevertheless much smaller than most light-water types).

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. The PBMR will have a direct-cycle gas turbine generator and thermal efficiency about 41%, the helium coolant leaving the bottom of the core at about 900°C. Up to 450,000 fuel pebbles 60 mm diameter and containing 9g uranium enriched to 10% uranium-235 recycle through the reactor continuously (about six times each, taking six months) until they are expended, giving an average enrichment in the fuel load of 5% and average burn-up of 80 GWday/t U (eventual target burn-ups are 200 GWd/t). This means on-line refuelling as expended pebbles (that have yielded up to 91 GWd/t) are replaced, giving high capacity factor. The reactor core 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%) without loss of thermal efficiency, and with rapid change in power settings. Power density in the core is about one-tenth of that in light water reactor, and if coolant circulation ceases, the fuel will survive initial high temperatures while the reactor shuts itself down—an inherent safety feature. Power is controlled by varying the coolant pressure and hence flow. Each unit will finally discharge about 19 tonnes/yr of spent pebbles to ventilated on-site storage bins.

The PBMR Demonstration Power Plant (DPP) is expected to start construction at Koeberg in 2009 and achieve criticality in 2013.  Eventual construction cost (when in clusters of four or eight units) is expected to be very competitive.  Investors in the PBMR project are Eskom, the South African Industrial Development Corporation and Westinghouse. The first commercial units are expected on line soon after the DPP and Eskom has said it expects to order 24, which justify fully commercial fuel supply and maintenance.  A contract for the pebble fuel plant at Pelindaba has been let.

Each 210g fuel pebble contains about 9g uranium and the total uranium in one fuel load is 4.1 t. MOX and thorium fuels are envisaged. With used fuel, the pebbles can be crushed and the 4% of their volume which is microspheres removed, allowing the graphite to be recycled. The company says microbial removal of carbon-14 is possible (also in the graphite reflectors when decommissioning).

A design certification application to the US Nuclear Regulatory Commission is expected.

A larger US design, the Gas Turbine-Modular Helium Reactor (MHR, or GT-MHR), will be built as modules of up to 600 MWt.  In its electrical application each would directly drive a gas turbine at 47% thermal efficiency, giving 280 MWe.  It can also be used for hydrogen production (100,000 t/yr claimed) and other high temperature process heat applications. The annular core consists of 102 hexagonal fuel element columns of graphite blocks with channels for helium coolant 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 up to 220 GWd/t, and coolant outlet temperature is 850°C with a target of 1000°C.

The MHR is being developed by General Atomics in partnership with Russia's OKBM, supported by Fuji (Japan) and Areva NP.  Initially it will be used to burn pure ex-weapons plutonium at Seversk (Tomsk) in Russia. A burnable poison such as Er-167 is needed for this fuel.  The preliminary design stage was completed in 2001, but the program to construct a prototype in Russia seems to have languished since.  Areva is working separately on a version of this called Antares.

The development timeline calls for a prototype to be constructed in Russia 2006-09 following regulatory review.

A smaller version of this design, the Remote-Site Modular Helium Reactor (RS-MHR) of 10-25 MWe, has been proposed by General Atomics. The fuel would be 20% enriched and refuelling interval would be 6-8 years.

A third full-size HTR design is Areva's Very High Temperature Reactor (VHTR) being put forward by Areva NP. It is based on the GT-MHR and has also involved Fuji. Reference design is 600 MW (thermal) with prismatic block fuel like the GT-MHR. Target core outlet temperature is 1000°C and it uses and indirect cycle, possibly with a helium-nitrogen mix in the secondary system. This removes the possibility of contaminating the generation or hydrogen production plant with radionuclides from the reactor core.

HTRs can potentially use thorium-based fuels, such as highly enriched uranium (HEU) with thorium, uranium-233 with thorium, and plutonium with thorium. Most of the experience with thorium fuels has been in HTRs. General Atomics say that the MHR has a neutron spectrum is such and the TRISO fuel so stable that the reactor can be powered fully with separated transuranic wastes (neptunium, plutonium, americium and curium) from light water reactor used fuel.  The fertile actinides enable reactivity control and very high burn-up can be achieved with it - over 500 GWd/t - the Deep Burn concept and hence DB-MHR design.  Over 95% of the Pu-239 and 60% of other actinides are destroyed in a single pass.

The three larger HTR designs, with the AHTR described below, are contenders for the US Next-Generation Nuclear Plant.

The Hyperion Power Module (HPM) is a small self-regulating hydrogen-moderated and potassium-cooled reactor producing 70 MWt /25 MWe fuelled by powdered uranium hydride.  It operates at about 550C and is designed to operate for 5 – 10 years before being returned to the factory for refuelling.  It is about 1.5 metres wide and 2 metres high, so easily portable, and has no moving parts.  Hyperion Power Generation has had preliminary discussions with the Nuclear Regulatory Commission and a US design certification application is possible in 2012, when the company plans to begin manufacturing the plants in New Mexico.  The design is licensed from the DOE Los Alamos laboratory there.  The company reported sales interest from Eastern Europe in August 2008, at $27 million per unit. The uranium hydride fuel in this is a powder and incorporates the hydrogen moderator. However, above 550C the UH3 of the fuel dissociates and the resulting decrease in moderator density reduces the core reactivity. As it cools, hydrogen is reabsorbed, increasing the core reactivity. All this is without much temperature change since the main energy gain or loss is involved in phase change. Outside the core is storage for hydrogen at a controlled temperature, which determines the hydrogen pressure in the whole system and the precise core temperature (increasing with higher pressure). The whole system thus becomes self-regulating, and is inherently safe. Enrichment level starts off at about 5% and over five years is burned down to about 3%, with criticality being maintained by steadily reducing the proportion of deuterium in the hydrogen. Another small US HTR concept is the Adams Atomic Engines 10 MWe direct simple Brayton cycle plant with low-pressure nitrogen as the reactor coolant and working fluid, and graphite moderation. The reactor core will be a fixed, annular bed with about 80,000 fuel elements each up to 60 mm diameter and containing approximately 9 grams of heavy metal as TRISO particles, with expected average burn-up of 80 GWd/t. The initial units will provide a reactor core outlet temperature of 800ºC and a thermal efficiency near 25%. Power output is controlled by limiting coolant flow. A demonstration plant is proposed for completion by 2011 with series production by 2014. The Adams Engine is deigned to be competitive with combustion gas turbines. ## Liquid Metal-cooled Fast Reactors Fast neutron reactors have no moderator, a higher neutron flux and are normally cooled by liquid metal such as sodium, lead, or lead-bismuth, with high conductivity and boiling point. They operate at or near atmospheric pressure and have passive safety features (most have convection circulating the primary coolant). Automatic load following is achieved due to the reactivity feedback—constrained coolant flow leads to higher core temperature which slows the reaction. Primary coolant flow is by convection. They typically use boron carbide control rods. The Encapsulated Nuclear Heat Source (ENHS) concept is a liquid metal-cooled reactor concept of 50 MWe being developed by the University of California. The core is in a metal-filled module sitting in a large pool of secondary molten metal coolant which also accommodates the eight separate and unconnected steam generators. There is convection circulation of primary coolant within the module and of secondary coolant outside it. Outside the secondary pool the plant is air cooled. Control rods would need to be adjusted every year or so and load-following would be autonomous. The whole reactor sits in a 17 metre deep silo. Fuel is a uranium-zirconium (U-Zr) alloy with 13% uranium enrichment (or uranium-plutonium-zirconium with 11% plutonium) with a 15-20 year life. After this, the module is removed, stored on site until the primary lead (or lead-bismuth) coolant solidifies, and it would then be shipped as a self-contained and shielded item. A new fuelled module would be supplied complete with primary coolant. The ENHS is designed for developing countries and is highly proliferation resistant but is not yet close to commercialization. A related project is the Secure Transportable Autonomous Reactor (STAR) being developed by Argonne National Laboratory under the leadership of Lawrence Livermore Laboratory of the U.S. Department of Energy (DOE). It a lead-cooled fast neutron modular reactor with passive safety features. Its 400 MWt size means it can be shipped by rail and cooled by natural circulation. It uses uranium-transuranic nitride fuel in a cassette which is replaced every 15-20 years. The STAR-LM was conceived for power generation, running at 578°C and producing 180 MWe. STAR-H2 is an adaptation for hydrogen production, with reactor heat at up to 800°C being conveyed by a helium circuit to drive a separate thermochemical hydrogen production plant, while lower-grade heat is harnessed for desalination (multi-stage flash process). Any commercial electricity generation then would be by fuel cells from the hydrogen. Its development is further off. A smaller STAR variant is the Small Sealed Transportable Autonomous Reactor (SSTAR), being developed in collaboration with Toshiba and others in Japan (see 4S below). It has lead or lead-bismuth cooling, runs at 566°C and has integral steam generator inside the sealed unit that would be installed below ground level. Conceived in sizes 10-100 MWe, main development is now focused on a 45 MWt/ 20 MWe version as part of the US Generation IV effort. After a 20-year life without refuelling, the whole reactor unit is then returned for recycling of the fuel. The core is one meter diameter and 0.8m high. SSTAR will eventually be coupled to a Brayton cycle turbine using supercritical carbon dioxide. Prototype envisaged for 2015. For all STAR concepts, regional fuel cycle support centers would handle fuel supply and reprocessing, and fresh fuel would be spiked with fission products to deter misuse. Complete burnup of uranium and transuranics is envisaged in STAR-H2, with only fission products being waste. Japan's LSPR is a lead-bismuth cooled reactor of 150 MWt /53 MWe. Fuelled units would be supplied from a factory and operate for 30 years, then be returned. This design concept is intended for developing countries. A small-scale design developed by Toshiba Corporation in cooperation with Japan's Central Research Institute of Electric Power Industry (CRIEPI) and funded by the Japan Atomic Energy Research Institute (JAERI) is the 5 MWt, 200 kWe Rapid-L, using lithium-6 (a liquid neutron poison) as a control medium. It would have 2700 fuel pins of 40-50% enriched uranium nitride with 2600°C melting point integrated into a disposable cartridge. The reactivity control system is passive, using lithium expansion modules (LEM) which give burnup compensation, partial load operation as well as negative reactivity feedback. As the reactor temperature rises, the lithium expands into the core, displacing an inert gas. Other kinds of lithium modules, also integrated into the fuel cartridge, shut down and start up the nuclear reactor. Cooling is by molten sodium, and with the LEM control system, reactor power is proportional to primary coolant flow rate. Refuelling would be every 10 years in an inert gas environment. Operation would require no skill, due to the inherent safety design features. The whole plant would be about 6.5 meters high and 2 meters in diameter. The Super-Safe, Small & Simple (4S) 'nuclear battery' system is being developed by Toshiba and CRIEPI in Japan in collaboration with STAR work and Westinghouse in USA. It uses sodium as coolant (with electromagnetic pumps) and has passive safety features, notably negative temperature and void reactivity. The whole unit would be factory-built, transported to site, installed below ground level, and would drive a steam cycle via a secondary sodium loop. It is capable of three decades of continuous operation without refuelling. Metallic fuel (169 pins 10mm diameter) is uranium-zirconium enriched to less than 20% or U-Pu-Zr alloy with 24% Pu for the 10 MWe version or 11.5% Pu for the 50 MWe version. Steady power output over the core lifetime is achieved by progressively moving upwards an annular reflector around the slender core (0.68m diameter, 2m high in the 10 MWe version, 1.2m diameter and 2.5m high in the 50 MWe version) at about one millimetre per week. The reflector height is slightly greater than the core height. After 14 years a neutron absorber at the centre of the core is removed and the reflector repeats its slow movement up the core for 16 more years. Burnup will be 34,000 MWday/t. In the event of power loss the reflector falls to the bottom of the reactor vessel, slowing the reaction, and external air circulation gives decay heat removal. A further safety device is a neutron absorber rod which can drop into the core. After 30 years the fuel would be allowed to cool for a year, then it would be removed and shipped for storage or disposal. Both 10 MWe and 50 MWe versions of 4S are designed to automatically maintain an outlet coolant temperature of 550°C—suitable for power generation with high temperature electrolytic hydrogen production. Plant cost is projected at US$2500/kW and power cost 5-7 cents/kWh for the small unit—very competitive with diesel in many locations. The design has gained considerable support in Alaska and toward the end of 2004 the town of Galena granted initial approval for Toshiba to build a 4S reactor in the remote location. A pre-application review by the Nuclear Regulatory Commission (NRC) is being sought with a view to application for design certification in 2009 and construction and operating licence (COL) application by 2012.   Its design is sufficiently similar to PRISM—GE's modular 150 MWe liquid metal-cooled inherently-safe reactor that went part-way through US NRC approval process, giving it favorable prospects for licensing. Toshiba plans a worldwide marketing program to sell the units for power generation at remote mines, desalination plants and for making hydrogen.  Eventually it expects sales for hydrogen production to outnumber those for power supply.

The L-4S is lead-bismuth cooled version of 4S.

A significant fast reactor prototype was the EBR-II, a fuel recycle reactor of 62 MWt at the U.S. Argonne National Laboratory (ANL) that used the pyrometallurgically-refined spent fuel from light water reactors as fuel, including a wide range of actinides. The objective of the program is to use the full energy potential of uranium rather then only about one percent of it. The reactor is now shut down and undergoing decommissioning. An EBR-III of 200-300 MWe was proposed but not developed.

Russia has experimented with several lead-cooled nuclear reactor designs, and has used lead-bismuth cooling for 40 years in its submarine reactors. lead-208 (54% of naturally-occurring lead) is transparent to neutrons. A significant Russian design is the BREST fast neutron reactor of 300 MWe or more with lead as the primary coolant, at 540°C, and supercritical steam generators. The core sits in a pool of lead at near atmospheric pressure. It is inherently safe and uses a uranium+plutonium nitride fuel. No weapons-grade plutonium can be produced (since there is no uranium blanket), and spent fuel can be recycled indefinitely at on-site facilities. A pilot unit was to be built at Beloyarsk and 1200 MWe units are planned.

A smaller and newer Russian design is the Lead-Bismuth Fast Reactor (SVBR) of 75-100 MWe from Gidropress.  This is an integral design, with the steam generators sitting in the same Pb-Bi pool at 400-480ºC as the reactor core.  It is designed to be able to use a wide variety of fuels, though the reference model uses uranium enriched to 16%.  Uranium-plutonium fuel is also envisaged.  The unit would be factory-made and shipped as a 4.5m diameter, 7.5m high module, then installed in a tank of water which gives passive heat removal and shielding.  A power station with 16 such modules is expected to supply electricity at lower cost than any other new Russian technology as well as achieving inherent safety and high proliferation resistance.  (Russia built 7 Alfa-class submarines, each powered by a compact 155 MWt Pb-Bi cooled reactor, essentially an SVBR, and 70 reactor-years operational experience was acquired with these.)  In mid 2008 Rosatom and Russkiye Mashiny (Russian Machines Co) put together a joint venture to build a civil SVBR reactor.  The plan is to complete the design development by 2017 and put on line a 100 MWe pilot facility by 2020, with total investment by Russkiye Mashiny of RUR16 billion ($585 million). The site selection process is underway – earlier plans were to put it Obninsk. The SVBR-100 could be the first reactor cooled by heavy metal to be utilized to generate electricity. In South Korea KAERI has been working on sodium-cooled fast reactor designs. A second stream of fast reactor development there is via the Nuclear Transmutation Energy Research Centre of Korea (NuTrECK) at Seoul University (SNU). It is working on a lead-bismuth cooled designs of 35 MW which would operate on pyro-processed fuel. It is designed to be leased for 20 years and operated without refuelling, then returned to the supplier. It would then be refuelled at the pyro-processing plant and have a design life of 60 years. It would operate at atmospheric pressure, eliminating major concern regarding loss of coolant accidents. ## Molten Salt Reactors (MSR) During the 1960s, the US developed the molten salt breeder reactor as the primary back-up option for the fast breeder reactor (FBR) (cooled by liquid metal) and a small prototype was operated at Oak Ridge National Laboratory (ORNL). There is now renewed interest in the concept in Japan, Russia, France and the USA, and one of the six generation IV designs selected for further development is the MSR. In the Molten Salt Reactor (MSR) the fuel is a molten mixture of lithium and beryllium fluoride salts with dissolved enriched uranium, thorium or uranium-233 fluorides. The core consists of an unclad graphite moderator arranged to allow the flow of salt at some 700°C and at low pressure. Heat is transferred to a secondary salt circuit and thence to steam. It is not a fast reactor, but with some moderation by the graphite is epithermal (intermediate neutron speed). The fission products dissolve in the salt and are removed continuously in an on-line reprocessing loop and replaced with thorium-232 or uranium-238. Actinides remain in the reactor until they fission or are converted to higher actinides which do so. A full-size 1000 MWe MSR breeder reactor was designed but not built. In 2002 a Thorium MSR was designed in France with a fissile zone where most power would be produced and a surrounding fertile zone where most conversion of Th-232 to U-233 would occur. The FUJI MSR is a 100 MWe design operating as a near-breeder and being developed internationally by a Japanese, Russian and US consortium. The attractive features of this MSR fuel cycle include: the high-level waste comprising fission products only, hence shorter-lived radioactivity; small inventory of weapons-fissile material (plutonium-242 being the dominant plutonium isotope); low fuel use (the French self-breeding variant claims 50kg of thorium and 50kg uranium-238 per billion kWh); and safety due to passive cooling up to any size. The Advanced High-temperature Reactor (AHTR) is a larger reactor using a coated-particle graphite-matrix fuel like that in the GTMHR (see above section) and with molten fluoride salt as primary coolant. While similar to the gas-cooled HTR, it operates at low pressure (less than 1 atmosphere) and higher temperature, and gives better heat transfer than helium. The salt is used solely as coolant, and achieves temperatures of 750-1000°C while at low pressure. This could be used in thermochemical hydrogen manufacture. Reactor sizes of 1000 MWe/2400 MWt are envisaged, with capital costs estimated at less than$1000/kW.

Molten fluoride salts are a preferred interface fluid between the nuclear heat source and any chemical plant. The aluminium smelting industry provides substantial experience in managing these materials safely. The hot molten salt can also be used with secondary helium coolant, generating power via the Brayton cycle.

## Modular construction

The IRIS developers have outlined the economic case for modular construction of their design (about 330 MWe), and the argument applies similarly to other smaller units.  They point out that IRIS with its size and simple design is ideally suited for modular construction.  The economy of scale is replaced here with the economy of serial production of many small and simple components and prefabricated sections.  They expect that construction of the first IRIS unit will be completed in three years, with subsequent reduction to only two years.

Site layouts have been developed with multiple single units or multiple twin units.  In each case, units will be constructed so that there is physical separation sufficient to allow construction of the next unit while the previous one is operating and generating revenue.  In spite of this separation, the plant footprint can be very compact so that a site with three IRIS single modules providing 1000 MWe is similar or smaller in size than one with a comparable total power single unit.

Eventually IRIS is expected to have a capital cost and production cost comparable with larger plants.  But any small unit such as this will potentially have a funding profile and flexibility otherwise impossible with larger plants.  As one module is finished and starts producing electricity, it will generate positive cash flow for the next module to be built. Westinghouse estimates that 1000 MWe delivered by three IRIS units built at three year intervals financed at 10% for ten years require a maximum negative cash flow less than \$700 million (compared with about three times that for a single 1000 MWe unit).  For developed countries small modular units offer the opportunity of building as necessary, for developing countries it may be the only option, because their electric grids cannot take 1000+ MWe single units.

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Glossary

### Citation

Hore-Lacy, I., & Association, W. (2010). Small nuclear power reactors. Retrieved from http://www.eoearth.org/view/article/156047