Small Nuclear Power Reactors
(Updated October 2014)
- There is revival of interest in small and simpler units for generating electricity from nuclear power, and for process heat.
- This interest in small and medium nuclear power reactors is driven both by a desire to reduce the impact of capital costs and to provide power away from large grid systems.
- The technologies involved are very diverse.
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 power reactors built for naval use (up to 190 MW thermal) and as neutron sourcesa, yielding enormous expertise in the engineering of small power units. The International Atomic Energy Agency (IAEA) defines 'small' as under 300 MWe, and up to about 700 MWe as 'medium' – including many operational units from 20th century. Together they are now referred to by IAEA as small and medium reactors (SMRs). However, 'SMR' is used more commonly as an acronym for 'small modular reactor', designed for serial construction and collectively to comprise a large nuclear power plant. (In this paper the use of diverse pre-fabricated modules to expedite the construction of a single large reactor is not relevant.) A subcategory of very small reactors – vSMRs – is proposed for units under about 15 MWe, especially for remote communities.
Today, due partly to the high capital cost of large power reactors generating electricity via the steam cycle and partly to the need to service small electricity grids under about 4 GWe,b there is a move to develop smaller units. These may be built independently or as modules in a larger complex, with capacity added incrementally as required (see section below on Modular construction using small reactor units). Economies of scale are provided by the numbers produced. There are also moves to develop independent small units for remote sites. Small units are seen as a much more manageable investment than big ones whose cost often rivals the capitalization of the utilities concerned.
This paper focuses on advanced designs in the small category, i.e. those now being built for the first time or still on the drawing board, and some larger ones which are outside the mainstream categories dealt with in the Advanced Reactors paper. Note that many of the designs described here are not yet actually taking shape. Three main options are being pursued: light water reactors, fast neutron reactors and also graphite-moderated high temperature reactors. The first has the lowest technological risk, but the second (FNR) can be smaller, simpler and with longer operation before refueling.
Generally, modern small reactors for power generation are expected to have greater simplicity of design, economy of mass production, and reduced siting costs. Most are also designed for a high level of passive or inherent safety in the event of malfunctionc. Also many are designed to be emplaced below ground level, giving a high resistance to terrorist threats. A 2010 report by a special committee convened by the American Nuclear Society showed that many safety provisions necessary, or at least prudent, in large reactors are not necessary in the small designs forthcomingd. Since small reactors are envisaged as replacing fossil fuel plants in many situations, the emergency planning zone required is designed to be no more than about 300 m radius.
A 2009 assessment by the IAEA under its Innovative Nuclear Power Reactors & Fuel Cycle (INPRO) program concluded that there could be 96 small modular reactors (SMRs) in operation around the world by 2030 in its 'high' case, and 43 units in the 'low' case, none of them in the USA. (In 2011 there were 125 small and medium units – up to 700 MWe – in operation and 17 under construction, in 28 countries, totaling 57 GWe capacity.) The IAEA has a program assessing a conceptual Multi-Application Small Light Water Reactor (MASLWR) design with integral steam generators, focused on natural circulation of coolant. The concept is similar to several of the integral PWR designs below.
A 2011 report for US DOE by University of Chicago Energy Policy Institute says development of small reactors can create an opportunity for the United States to recapture a slice of the nuclear technology market that has eroded over the last several decades as companies in other countries have expanded into full‐scale reactors for domestic and export purposes. However, it points out that detailed engineering data for most small reactor designs are only 10 to 20 percent complete, only limited cost data are available, and no US factory has advanced beyond the planning stages. In general, however, the report says small reactors could significantly mitigate the financial risk associated with full‐scale plants, potentially allowing small reactors to compete effectively with other energy sources.
In January 2012 the DOE called for applications from industry to support the development of one or two US light-water reactor designs, allocating $452 million over five years. Four applications were made, from Westinghouse, Babcock & Wilcox, Holtec, and NuScale Power, the units ranging from 225 down to 45 MWe. DOE announced its decision in November 2012 to support the B&W 180 MWe mPower design, to be developed with Bechtel and TVA. Through the five-year cost-share agreement, the DOE will invest up to half of the total project cost, with the project's industry partners at least matching this. The total will be negotiated between DOE and B&W, up to $226 million.
In March 2013 the DOE called for applications for second-round funding, and proposals were made by Westinghouse, Holtec, NuScale, General Atomics, and Hybrid Power Technologies, the last two being for EM2 and Hybrid SMR, not PWRs. Other (non-PWR) small reactor designs will have modest support through the Reactor Concepts RD&D program. A late application ‘from left field’ was from National Project Management Corporation (NPMC) which includes a cluster of regional partners in the state of New York, South Africa’s PBMR company, and National Grid, the UK-based grid operator with 3.3 million customers in New York, Massachusetts and Rhode Island. The project is for a HTR of 165 MWe, apparently the earlier direct-cycle version of the shelved PBMR, emphasising its ‘deep burn’ attributes in destroying actinides and achieving high burn-up at high temperatures. The PBMR design was a contender with Westinghouse backing for the US Next-Generation Nuclear Power (NGNP) project, which has stalled since about 2010.
In December 2013 DOE announced that a further grant would be made to NuScale on a 50-50 cost-share basis, to support design development and NRC certification and licensing of its 45 MWe small reactor design.
In March 2012 the US DOE signed agreements with three companies interested in constructing demonstration small reactors at its Savannah River site in South Carolina. The three companies and reactors are: Hyperion with a 25 MWe fast reactor, Holtec with a 140 MWe PWR, and NuScale with 45 MWe PWR. DOE is discussing similar arrangements with four further small reactor developers, aiming to have in 10-15 years a suite of small reactors providing power for the DOE complex. DOE is committing land but not finance. (Over 1953-1991, Savannah River was where a number of production reactors for weapons plutonium and tritium were built and run.)
In January 2014 Westinghouse announced that was suspending work on its small modular reactors in the light of inadequate prospects for multiple deployment. The company said that it could not justify the economics of its SMR without government subsidies, unless it could supply 30 to 50 of them. It was therefore delaying its plans, though small reactors remain on its agenda.
The most advanced small modular reactor project is in China, where Chinergy is starting to build the 210 MWe HTR-PM, which consists of twin 250 MWt high-temperature gas-cooled reactors (HTRs) which build on the experience of several innovative reactors in the 1960s to1980s.
Another significant line of development is in very small fast reactors of under 50 MWe. Some are conceived for areas away from transmission grids and with small loads; others are designed to operate in clusters in competition with large units.
Urenco has called for European development of very small – 5 to 10 MWe – 'plug and play' inherently-safe reactors based on graphite-moderated HTR concepts. It is seeking government support for a prototype "U-Battery" which would run for 5-10 years before requiring refueling or servicing.
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 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.
Also in the small reactor category are the Indian 220 MWe pressurised heavy water reactors (PHWRs) based on Canadian technology, and the Chinese 300-325 MWe PWR such as built at Qinshan Phase I and at Chashma in Pakistan, and now called CNP-300. The Nuclear Power Corporation of India (NPCIL) is now focusing on 540 MWe and 700 MWe versions of its PHWR, and is offering both 220 and 540 MWe versions internationally. These small established designs are relevant to situations requiring small to medium units, though they are not state of the art technology.
Other, mostly larger new designs are described in the information page on Advanced Nuclear Power Reactors.
Small (25 MWe up) reactors operating
||CNNC, operational in Pakistan
Small (25 MWe up) reactor designs under construction
||CNEA & INVAP, Argentina
||INET & Huaneng, China
Small (25 MWe up) reactors for near-term deployment – development well advanced
||Babcock & Wilcox + Bechtel, USA
||CNNC & Guodian, China
||KAERI, South Korea
||NuScale Power + Fluor, USA
||PBMR, South Africa; NPMC, USA*
Small (25 MWe up) reactor designs at earlier stages
||General Atomics (USA)
||Gen4 (Hyperion), USA
||5, 100 MWe
||29, 120, 288 MWe
||Terrestrial Energy, Canada
||Northern Nuclear, Canada
* well-advanced designs understood to be on hold
Light water reactors
These are moderated and cooled by ordinary water and have the lowest technological risk, being similar to most operating power and naval reactors today. They mostly use fuel enriched to less than 5% U-235 with no more than six-year refueling interval, and regulatory hurdles are likely least of any small reactors.
US experience of small light water reactors (LWRs) has been of very small military power plants, such as the 11 MWt, 1.5 MWe (net) PM-3A reactor which operated at McMurdo Sound in Antarctica 1962-72, generating a total of 78 million kWh. It was refueled once, in 1970. There was also an Army program for small reactor development, most recently the DEER (deployable electric energy reactor) concept which was being commercialised by Radix Power & Energy. DEER would be portable and sealed, able to operate in the range 3 to 10 MWe, for forward military bases. 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.
The US Nuclear Regulatory Commission is starting to focus on small light-water reactors using conventional fuel, such as B&W, Westinghouse, NuScale, and Holtec designs including integral types (B&W, Westinghouse, NuScale). Beyond these in time and scope, “the NRC intends to take full advantage of the experience and expertise” of other nations which have moved forward with non light-water designs, and it envisages “having a key role in future international regulatory initiatives.”
Of the following designs, the KLT, VBER and Holtec SMR have conventional pressure vessels 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 LWRs. All require conventional cooling of the steam condenser.
In the USA major engineering and construction companies have taken active shares in two projects: Fluor in NuScale, and Bechtel in B&W mPower.
Three new concepts are alternatives to conventional land-based nuclear power plants. Russia's floating nuclear power plant (FNPP) with a pair of PWRs derived from icebreakers is well on the way to commissioning, with the KLT-40S reactors described below and in the Nuclear Power in Russia paper. France's submerged Flexblue power plant, using a 50-250 MWe reactor, probably NP-300 described below, is an early concept, as is MIT’s floating plant moored offshore with a reactor of about 200 MWe in the bottom part of a cylindrical platform.
Russia's KLT-40S from OKBM Afrikantov is a reactor well proven in icebreakers and now – with low-enriched fuel – 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 with on-board refuelling capability and used 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 used fuel. Two units will be mounted on a 20,000 tonne barge to allow for outages (70% capacity factor). It may also be used in Kaliningrad.
Although the reactor core is normally cooled by forced circulation (four-loop), the design relies on convection for emergency cooling. Fuel is uranium aluminium silicide with enrichment levels of up to 20%, giving up to four-year refuelling intervals. A variant of this is the KLT-20, specifically designed for FNPP. It is a 2-loop version with same enrichment but 10-year refueling interval.
The first floating nuclear power plant, the Akademik Lomonosov, commenced construction in 2007. Due to insolvency of the shipyard the plant is now expected to be completed in late 2016.2 (See also Floating nuclear power plants section in the information page on Nuclear Power in Russia.)
OKBM Afrikantov is developing a new compact icebreaker reactor – RITM-200 – to replace the KLT reactors and to serve in floating nuclear power plants. This is an integral 175 MWt, 50 MWe PWR (also quoted at 210 MWt, 55 MWe) with 4 coolant loops and external main circulation pumps. It has inherent safety features, using low-enriched (<20%) fuel. Refueling is every seven years at 65% capacity factor, over a 40-year total lifespan. It is designed to provide 30 MW shaft power for an icebreaker, and the LK-60 design will be powered by two of them. The reactor plant in containment has a mass of 1100 tonnes and is 6 m × 6 m × 15.5 m. A major challenge is the reliability of steam generators and associated equipment which are much less accessible when inside the reactor pressure vessel.
This is based on the Qinshan 1 reactor in China as a two-loop PWR operating in Pakistan and with further units being built there. It is 1000 MWt, 325 MWe with design life 40 years. Fuel enrichment is 2.4-3.0%, fuel cycle 12 months. It is from China National Nuclear Corp (CNNC).
A smaller unit is the NuScale multi-application small reactor, a 160 MWt, 50 MWe integral PWR with natural circulation. In December 2013 the US Department of Energy (DOE) announced that it would support accelerated development of the design for early deployment on a 50-50 cost share basis. An agreement for $217 million over five years was signed in May 2014 by NuScale Power.
It will be factory-built with 3-metre diameter pressure vessel and convection cooling, with the only moving parts being the control rod drives. It uses standard PWR fuel enriched to 4.95% in normal PWR fuel assemblies (but which are only 2 m long), with 24-month refuelling cycle. Installed in a water-filled pool below ground level, the 4.6 m diameter, 22 m high cylindrical containment vessel module weighs 650 tonnes and contains the reactor with steam generator above it. A standard power plant would have 12 modules together giving about 600 MWe. An overhead crane would hoist each module from its pool to a separate part of the plant for refueling. Design life is 60 years. It has full passive cooling in operation and after shutdown for an indefinite period.
The NuScale Power company was spun out of Oregon Sate University in 2007, though the original development was funded by the US Department of Energy. The company estimated in 2010 that overnight capital cost for a 12-module, 540 MWe NuScale plant would be about $4000 per kilowatt. After NuScale experienced problems in funding its development, Fluor Corporation paid over $30 million for 55% of NuScale in October 2011. With the support of Fluor, NuScale expects to bring its technology to market in a timely manner. The DOE sees this as a "near-term LWR design." In August 2013 Rolls Royce joined the venture to support an application for DOE funding, and in March 2014 Enercon Services took undisclosed equity to become a partner and assist with design certification and COL applications.
NuScale expects to lodge an application for US design certification late in 2016, and is already engaged with NRC, having spent some $130 million on licensing to November 2013. It expects the NRC review to take 39 months, so the first unit could be under construction in 2020.
In March 2012 the US DOE signed an agreement with NuScale regarding constructing a demonstration unit at its Savannah River site in South Carolina.
In mid-2013 NuScale launched the Western Initiative for Nuclear (WIN) – a broad, multi-western state collaboration* – to study the demonstration and deployment of a multi-module NuScale Small Modular Reactor (SMR) plant in the western USA. A NuScale SMR built as part of Project WIN is projected to be operational by 2024, likely in Idaho, followed by a second in Washington state. WIN includes Energy Northwest (ENW) in Washington and Utah Associated Municipal Power Systems (UAMPS). In mid-2014 the plan was for UAMPS to be the owner and ENW the operator of a plant built at DOE’s Idaho National Laboratory site. The UAMPS Carbon-Free Power Project will comprise a 540-600 MWe power plant (12 modules), costing $5000/kW on overnight basis, hence about $3.0 billion. Energy Northwest comprises 27 public utilities, and has examined small reactor possibilities before choosing NuScale.
* Washington, Oregon, Idaho, Wyoming, Utah and Arizona.
In mid-2009, Babcock & Wilcox (B&W) announced its B&W mPower reactor, a 500 MWt, 180 MWe integral PWR designed to be factory-made and railed to sitei. In November 2012 the US Department of Energy (DOE) announced that it would support accelerated development of the design for early deployment, with up to $226 million.
The reactor pressure vessel containing core of 2x2 metres and steam generator is thus only 3.6 metres diameter and 22 m high, and the whole unit 4.5 m diameter and 23 m high. It would be installed below ground level, have an air-cooled condenser giving 31% thermal efficiencyp, and passive safety systems. The power was originally 125 MWe, but as of mid-2012, 180 MWe is quoted when water-cooled. A 155 MWe air-cooled version is also planned. The integral steam generator is derived from marine designs, as is the control rod set-up. It has a "conventional core and standard fuel" (69 fuel assemblies, each standard 17x17, < 20 t)j enriched to almost 5%, with burnable poisons, to give a four-year operating cycle between refuelling, which will involve replacing the entire core as a single cartridge. Core power density is lower than in a large PWR, and burn-up is about 35 GWd/t. (B&W draws upon over 50 years experience in manufacturing nuclear propulsion systems for the US Navy, involving compact reactors with long core life.) A 60-year service life is envisaged, as sufficient used fuel storage would be built on site for this.
The mPower reactor is modular in the sense that each unit is a factory-made module and several units would be combined into a power station of any size, but most likely 360-720 MWe (2, 3 or 4 units) and using 250 MWe turbine generators (also shipped as complete modules), constructed in three years. B&W's present manufacturing capability in North America can produce these units, and B&W Nuclear Energy Inc set up B&W Modular Nuclear Energy LLC (B&W MNE) to market the design, in collaboration with Bechtel which joined the project as an equity partner to design, licence and deploy it. B&W's 90%-owned subsidiary, Generation mPower LLC (GmP), reports into B&W MNE. B&W expected both design certification and construction permit in 2018, and commercial operation of the first two units in 2022. Meanwhile the design is phase 1 of the Canadian Nuclear Safety Commission licensing process.
In November 2013 B&W said it would seek to bring in further equity partners by mid-2014 to take forward the licensing and construction of an initial plant.* B&W said it had invested $360 million in GmP with Bechtel, and wanted to sell up to 70% of its stake in the JV, leaving it with about 20% and Bechtel 10%. In April 2014 B&W announced that it was cutting back funding on project to about $15 million per year, having failed to find customers or investors. This will necessitate some renegotiation with DOE in respect to funding from that quarter, though about $101 million has already been paid. B&W planned to retain the rights to manufacture the reactor module and nuclear fuel for the mPower plant. In August 2014 the TVA said it would file an early site permit (ESP) application instead of a construction permit application for one or more small modular reactors at Clinch River, possibly by the end of 2015.
Overnight cost for a twin-unit plant is put by B&W at about $5000/kW.
Westinghouse's IRIS (International Reactor Innovative & Secure) is an advanced reactor design which has evolved over more than two decades. A 1000 MWt, 335 MWe capacity was proposed, although it could be scaled down to 100 MWe. IRIS is a modular pressurised water reactor with integral primary coolant system and circulation by convection. Fuel is similar to present LWRs and (at least for the 335 MWe version) fuel assemblies would be identical to those in AP1000. Enrichment is 5% with burnable poison and fuelling interval of up to four years (or longer with higher enrichment and MOX fuel). US design certification was at the pre-application stage, but the concept appears to have evolved into the Westinghouse SMR.
This Small Modular Reactor is an 800 MWt/ 225 MWe class integral PWR with passive safety systems and reactor internals including fuel assemblies based closely on those in the AP1000 (89 assemblies 2.44m active length, <5% enrichment). The steam generator is above the core fed by 8 horizontally-mounted axial-flow coolant pumps. The reactor vessel will be factory-made and shipped to site by rail, then installed below ground level in a containment vessel 9.8 m diameter and 27 m high. The reactor vessel module is 25 metres high and 3.5 metres diameter. It has a 24-month refueling cycle and 60-year service life. Passive safety means no operator intervention is required for 7 days in the event of an accident. In May 2012 Westinghouse teamed up with General Dynamics Electric Boat to assist in the design and Burns & McDonnell to provide architectural and engineering support. A design certification application was expected by NRC in September 2013, but the company has stepped back from lodging one while it re-assesses the market for small reactors. The company has started fabricating prototype fuel assemblies.
The DOE sees this as a "near-term LWR design." In April 2012 Westinghouse set up a project with Ameren Missouri to seek DOE funds for developing the design, with a view to obtaining design certification and a combined construction and operation licence (COL) from the Nuclear Regulatory Commission for up to five SMRs at Ameren's Callaway site, instead of an earlier proposed large EPR there. The initiative – NexStart SMR Alliance – had the support of other state utilities and the state governor, as well as Savannah River, Exelon and Dominion. However, this agreement expired about the end of 2013, and both companies stepped back from the project as DOE funds went to other SMR projects.
In May 2013 Westinghouse announced that it would work with China’s State Nuclear Power Technology Corporation (SNPTC) to accelerate design development and licensing in the USA and China of its SMR. SNPTC would ensure that the Westinghouse SMR design met standards for licensing in China and would lead the licensing effort in that country. The status of this collaboration is uncertain.
Holtec International set up a subsidiary – SMR LLC – to commercialize a 140 MWe (446 MWt) factory-built reactor concept called Holtec Inherently Safe Modular Underground Reactor (HI-SMUR). The particular design being promoted is a 160 MWe version of this, SMR-160, with two external horizontal steam generators, using fuel similar to that in larger PWRs, including MOX. The 32 full-length fuel assemblies are in a fuel cartridge, which is loaded and unloaded as a single unit from the 31-metre high pressure vessel. Holtec claims a one-week refueling outage every 42 months. It has full passive cooling in operation and after shutdown for an indefinite period, and also a negative temperature coefficient so that it shuts down at high temperatures. The reactor will be offered with optional heat sink to atmosphere, using dry cooling. The whole reactor system will be installed below ground level, with used fuel storage. A 24-month construction period is envisaged for each $800 million unit ($5000/kW). Operational life claimed is 80 years.
Holtec expects to submit an application for design certification to NRC late in 2016. The detailed design phase is from August 2012, and it is apparently not as far ahead as the other three US small designs. The Shaw Group (CB&I subsidiary) is providing engineering support for the design, and in June 2013 URS Corporation joined to support design and qualification. Holtec expects its involvement to take a year off the development schedule. The Construction Permit Application and Preliminary Safety Analysis Report are due in June 2014.
In March 2012 the US DOE signed an agreement with Holtec regarding constructing a demonstration SMR-160 unit at its Savannah River site in South Carolina. NuHub, a South Carolina economic development project, and the state itself supported Holtec's bid for DOE funding for the SMR-160, as did partners PSEG and SCE&G – which would operate the demonstration plant. Exelon, Entergy and FirstEnergy (though see above re mPower) were also supporters of the bid. Apart from the SCE&G demonstration plant, Holtec was negotiating to supply a SMR-160 to PSEG for its Hope Creek/Salem site in New Jersey, for which PSEG has sought an early site permit (ESP). After failing to get DOE funding, both PSEG and SCE&G reaffirmed their support for the SMR-160.
This is a 850 MWt, 300 MWe 2-loop PWR design from Gidropress, based on the VVER-640 (V-407) design. It is little reported.
A larger Russian factory-built and barge-mounted unit (requiring a 12,000 tonne vessel) is the VBER-150, of 350 MWt, 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 burn-up (31 GWd/t average, 41.6 GWd/t maximum) and eight-year refuelling interval.
OKBM Afrikantov's larger VBER-300 PWR is a 917 MWt, 295-325 MWe unit, the first of which is planned to 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 external steam generators and a cassette core with 85 standard VVER 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 two years out to five years (6 to 15 years fuel cycle) with burn-up to 125 GWd/tU. A 2006 joint venture between Atomstroyexport and Kazatomprom set this up for development as a basic power source in Kazakhstan, then for exporte. It is also envisaged for use in Russia, mainly as cogeneration unit. It is considered likely for near-term deployment.
The company also offers 200-600 MWe designs based on a standard 100 MWe module and explicitly based on naval units.
Another larger Russian reactor at the conceptual design stage is the VK-300 boiling water reactor of 750 MWt being developed specifically for cogeneration of both power and district heating or heat for desalination (150 MWe plus 1675 GJ/hr) by the N.A. Dollezhal Research and Development Institute of Power Engineering (RDIPE or NIKIET) together with several major research and engineering institutes. It has evolved from the 50 MWe (net) VK-50 BWR at Dimitrovgradf, but uses standard components wherever possible, and fuel elements similar to the VVER. Cooling is passive, by convection, and all safety systems are passive. Fuel enrichment is 4% and burn-up is 41 GWd/tU with 18-month refueling. It is capable of producing 250 MWe if solely electrical. In September 2007 it was announced that six would be built at Kola or Archangelsk and at Primorskaya in the far east, to start operating 2017-20,4 but no more has been heard of this plan. As a cogeneration plant it was intended for the Mining & Chemical Combine at Zheleznogorsk, but MCC is reported to prefer the VBER-300.
A smaller Russian BWR design is the 12 MWe transportable VKT-12, described as similar to the VK-50 prototype BWR at Dimitrovgrad, with one loop. It has a ceramic-metal core with uranium enriched to 2.4-4.8%, and 10-year refuelling interval. The reactor vessel is 2.4m inside diameter and 4.9 m high.
A smaller Russian OKBM Afrikantov PWR unit under development is the ABV, with a range of sizes from 45 MWt (ABV-6M ) down to 18 MWt (ABV-3), giving 4-18 MWe outputs. (The IAEA 2011 write-up quotes 45 MWt and 8.6 MWe in condensation mode and 14 MWt and 6 MWe in cogeneration mode.) The units are compact, with integral steam generator and natural circulation in the primary circuit. The units will be factory-produced and designed as a universal power source for floating NPPs – the ABV-6M would require a 3500 tonne barge; the ABV-3, 1600 tonne for twin units. The land-based version has reactor module 13 m long and 8.5m diameter, with mass 600 t. The core is similar to that of the KLT-40 except that enrichment is 16.5% or 19.7% and average burn-up 95 GWd/t. It would initially be fuelled in the factory. Refuelling interval is about 8-12 years, and service life about 60 years.
The CAREM-25 reactor prototype being built by the Argentine National Atomic Energy Commission (CNEA), with considerable input from INVAPg, is an older design modular 100 MWt (27 MWe gross) pressurised water reactor, first announced in 1984. It has 12 integral steam generators and is designed to be used for electricity generation 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 (11m high, 3.5m diameter), self-pressurised and relying entirely on convection (for modules less than 150 MWe). The final full-sized export version will be about 300 MWe, with axial coolant pumps driven electrically. Fuel is standard 3.1 or 3.4% enriched PWR fuel in hexagonal fuel assemblies, with burnable poison, and is refuelled annually.
The 25 MWe prototype unit is being built next to Atucha, on the Parana River in Lima, 110 km northwest of Buenos Aires, and the first larger version (probably 100 MWe) is planned in the northern Formosa province, 500 km north of Buenos Aries, once the design is proven. Some 70% of CAREM-25 components will be local manufacture. The IAEA lists it as a research reactor under construction since April 2013, though first concrete was poured in February 2014, marking official start of construction.
On a larger scale, South Korea's SMART (System-integrated Modular Advanced Reactor) is a 330 MWt pressurised water reactor with integral steam generators and advanced safety features. It is designed by the Korea Atomic Energy Research Institute (KAERI) for generating electricity (up to 100 MWe) and/or thermal applications such as seawater desalination. Design life is 60 years, fuel enrichment 4.8%, with a three-year refuelling cycle. Residual heat removal is passive. While the basic design is complete, the absence of any orders for an initial reference unit has stalled development. It received standard design approval from the Korean regulator in mid 2012 and KAERI plans to build a 90 MWe demonstration plant to operate from 2017. A single unit can produce 90 MWe plus 40,000 m3/day of desalinated water.
The Japan Atomic Energy Research Institute (JAERI) designed 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. Little has been heard of it since the start of the Millennium.
Technicatome (Areva TA) in France has developed the NP-300 PWR design from submarine power plants and aimed it at export markets for power, heat and desalination. It has passive safety systems and could be built for applications of 100 to 300 MWe or more with up to 500,000 m3/day desalination. Areva TA makes the K15 naval reactor of 150 MW, running on low-enriched fuel, and the land-based equivalent: Réacteur d’essais à terre (RES) a test version of which is under construction at Cadarache, due to operate about 2011.
It appears that some version of this reactor will be used in the Flexblue submerged nuclear power plant being proposed by DCNS in France. DCNS considered starting to build a prototype Flexblue unit in 2013 in its shipyard at Cherbourg for launch and deployment in 2016. The concept eliminates the need for civil engineering, and refuelling or major service can be undertaken by refloating it and returning to the shipyard.
The Chinese NHR-200 (Nuclear Heating Reactor), developed by Tsingua University's Institute of Nuclear Energy Technology (now the Institute of Nuclear and New Energy Technology), is a simple 200 MWt integral PWR design for district heating or desalination. It is based on the NHR-5 which was commissioned in 1989, and runs at lower temperature than the above designsh. Used fuel is stored around the core in the pressure vessel. In 2008, the Chinese government was reported to have agreed to build a multi-effect distillation (MED) desalination plant using this on the Shandong peninsula, but no more has been heard of it, and INET is focused on the HTR-PM being built in Shandong.
China National Nuclear Corporation (CNNC) has designed a multi-purpose small modular reactor, the ACP100. This is being developed by China New Energy Co Ltd. It has passive safety features, notably decay heat removal, and will be installed underground. It has 57 fuel assemblies 2.15m tall and integral steam generators (287°C), so that the whole steam supply system is produced and shipped a single reactor module. Its 310 MWt produces about 100 MWe, and power plants comprising two to six of these are envisaged, with 60-year design life and 24-month refueling. Or each module can supply 1000 GJ/hr, giving 12,000 m3/day desalination (with MED). Industrial and district heat uses are also envisaged, as well as floating nuclear power plant (FNPP) applications. Capacity of up to 150 MWe is possible.
CNNC New Energy Corporation, a joint venture of CNNC (51%) and China Guodian Corp, is planning to build two ACP100 units in Putian county, Zhangzhou city, at the south of Fujian province, near Xiamen, as a demonstration plant. This will be the CNY 5 billion ($788 million) phase 1 of a larger project. Completion of preliminary design is expected in 2014, with construction start in 2015 and operation in 2017. Construction time is expected to be 36-40 months. It involves a joint venture of three companies for the pilot plant: CNNC as owner and operator, the Nuclear Power Institute of China (NPIC) as the reactor designer and China Nuclear Engineering Group being responsible for plant construction.
The company signed a second ACP100 agreement with Hengfeng county, Shangrao city in Jiangxi province, and a third with Ningdu county, Ganzhou city in Jiangxi province in July 2013 for another ACP100 project costing CNY 16 billion. Further inland units are planned in Hunan and possibly Jilin provinces. Export potential is considered to be high, with full IP rights.
This is an integral PWR, with SNPTC provenance, being developed from the CAP1000 in parallel with CAP1400 by SNERDI, using proven fuel and core design. It is 450 MWt/150 MWe and has eight integral steam generators (295°C), and claims “a more simplified system and more safety than current third generation reactors”. It is pitched for remote electricity supply and district heating, with three-year refueling and design life of 80 years. It has both active and passive cooling and in an accident scenario, no operator intervention required for seven days. Seismic design basis 300 Gal. In mid-2013 SNPTC quoted approx. $5000/kW capital cost and 9 c/kWh, so significantly more than the CAP1400.
In China, a SNERDI project is a reactor for floating nuclear power plant (FNPP). This is to be 200 MWt and relatively low-temperature (250°C), so only about 40 MWe with two external steam generators and five-year refueling.
China General Nuclear Group (CGN) has two small ACPR designs: an ACPR100 and ACPR50S, both with passive cooling for decay heat and 60-year design life. Both have standard type fuel assemblies and fuel enriched to <5% with burnable poison giving 30-month refueling. The ACPR100 is an integral PWR, 450 MWt, 140 MWe, having 69 fuel assemblies. Reactor pressure vessel is 17m high and 4.4 m inside diameter, operating at 310°C. It is designed as a module in larger plant and would be installed underground. The offshore ACPR50S is 200 MWt, 60 MWe with 37 fuel assemblies and four external steam generators. Reactor pressure vessel is 7.4m high and 2.5 m inside diameter, operating at 310°C. It is designed for mounting on a barge as floating nuclear power plant (FNPP). The applications for these are similar to those for the ACP100, but the timescale is longer.
This is a conceptual design from DCNS (a state-owned defence group), Areva, EdF and CEA from France. It is designed to be submerged, 60-100 metres deep on the sea bed up to 15 km offshore, and returned to a dry dock for servicing. The reactor, steam generators and turbine-generator would be housed in a submerged 12,000 tonne cylindrical hull about 100 metres long and 12-15 metres diameter. Each hull and power plant would be transportable using a purpose-built vessel. Reactor capacity is 50-250 MWe, derived from DCNS's latest naval designs, but with details not announced. When first announced early in 2011 it was said that DCNS could start building a prototype Flexblue unit in 2013 in its shipyard at Cherbourg for launch and deployment in 2016, possibly off Flamanville.
This is an integral 5-10 MWe PWR conceptual design from Russia’s Research and Development Institute of Power Engineering (RDIPE). A 20 MWt version has three coolant loops, with natural circulation, and claims self-regulation with burnable poisons in unusual metal-ceramic fuel design, so needs no more than an annual maintenance campaign and no refueling during a 25-year life. The mass of one unit with shielding is 180 tonnes, so it can be shipped complete from the factory to site.
This is a Russian 6 MWe PWR concept with turbogenerator in a cylindrical pod about 15 m long and 8 m diameter, sitting on the sea bed like Flexblue. The reactor is based on operating prototypes, and would be serviced infrequently. It is intended as energy supply for oil and gas developments in Arctic seas.
Mitsubishi Heavy Industries have a conceptual design of Integrated Modular Reactor (IMR), a PWR of 1000 MWt, 350 MWe. It has design life of 60 years, 4.8% fuel enrichment and fuel cycle of 26 months. It has natural circulation for cooling. The project has involved Kyoto University, the Central Research Institute of the Electric Power Industry (CRIEPI), and the Japan Atomic Power Company (JAPC), with funding from METI. The target year to start licensing is 2020 at the earliest.
The TRIGA Power System is a 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 perfluorocarbon. The fuel is uranium-zirconium hydride enriched to 20% and with a little burnable poison and requiring refuelling every 18 months. Used fuel is stored inside the reactor vessel.
The Fixed Bed Nuclear Reactor (FNBR) is an early conceptual design from the Federal University of Rio Grande do Sul, Brazil. It a PWR with pebble fuel, 134 MWt, 70 MWe, with “flexible fuel cycle”.
Heavy water reactors
These are the oldest and smallest of the Indian pressurized heavy water reactor (PHWR) range, with a total of 16 now on line, 800 MWt, 220 MWe gross typically. Rajasthan 1 was built as a collaborative venture between Atomic Energy of Canada Ltd (AECL) and the Nuclear Power Corporation of India (NPCIL), starting up in 1972. Subsequent indigenous PHWR development has been based on these units, though several stages of evolution can be identified: PHWRs with dousing and single containment at Rajasthan 1-2, PHWRs with suppression pool and partial double containment at Madras, and later standardized PHWRs from Narora onwards having double containment, suppression pool, and calandria filled with heavy water, housed in a water-filled calandria vault. They are moderated and cooled by heavy water, and the natural uranium oxide fuel is in horizontal pressure tubes, allowing refueling on line (maintenance outages are scheduled after 24 months). Burn-up is about 15 GWd/t.
The Advanced Heavy Water Reactor developed by the Bhaba Atomic Research Centre (BARC) is designed to make extensive use of India’s abundant thorium as fuel, but a low-enriched uranium fuelled version is pitched for export. This will use low-enriched uranium plus thorium as a fuel, largely dispensing with the plutonium input of the version for domestic use. About 39% of the power will come from thorium (via in situ conversion to U-233, cf two thirds in domestic AHWR), 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. It will have vertical pressure tubes in which the light water coolant under high pressure will boil, circulation being by convection. It is at basic design stage.
High-temperature gas-cooled reactors
These use graphite as moderator (unless fast neutron type) and either helium, carbon dioxide or nitrogen as primary coolant. The experience of several innovative reactors built in the 1960s and 1970sk has been analysed, especially in the light of US plans for its Next Generation Nuclear Plant (NGNP) and China's launching its HTR-PM project in 2011. Lessons learned and documented for NGNP include the use of TRISO fuel, use of a reactor pressure vessel, and use of helium cooling (UK AGRs are the only HTRs to use CO2 as primary coolant). However US government funding for NGNP has now virtually ceased.
New high-temperature gas-cooled reactors (HTRs) are being developed which will be capable of delivering high temperature (700-950ºC and eventually up to about 1000°C) helium either for industrial application via a heat exchanger, or to make steam conventionally in a secondary circuit via a steam generator, or directly to drive a Brayton cycle* gas turbine for electricity with almost 50% thermal efficiency possible (efficiency increases around 1.5% with each 50°C increment). One uses the helium to drive an air compressor to supercharge a CCGT unit. Improved metallurgy and 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. All but one of those described below have neutron moderation by graphite, one is a fast neutron reactor.
* There is little interest in pursuing direct Brayton cycle for primary helium at present due to high technological risk.
Fuel for these reactors is in the form of TRISO (tristructural-isotropic) particles less than a millimetre in diameter. Each has a kernel (ca. 0.5 mm) of uranium oxycarbide (or uranium dioxide), with the uranium enriched up to 20% U-235, though normally less. This is surrounded by layers of carbon and silicon carbide, giving a containment for fission products which is stable to over 1600°C.
There are two ways in which these 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 used fuel than from the same capacity in a light water reactor. The moderator is graphite.
HTRs can potentially use thorium-based fuels, such as highly-enriched or low-enriched uranium with Th, U-233 with Th, and Pu with Th. Most of the experience with thorium fuels has been in HTRs (see information paper on Thorium).
With negative temperature coefficient of reactivity (the fission reaction slows as temperature increases) and passive decay heat removal, the reactors are inherently safe. HTRs therefore do not require any containment building for safety. They are sufficiently small to allow factory fabrication, and will usually be installed below ground level.
Three HTR designs in particular – PBMR, GT-MHR and Antares/ SC-HTGR – were contenders for the Next Generation Nuclear Plant (NGNP) project in the USA (see Next Generation Nuclear Plant section in the information page on US Nuclear Power Policy). In 2012 Antares was chosen. However, the only HTR project currently proceeding is the Chinese HTR-PM.
Hybrid Power Technologies have a hybrid-nuclear Small Modular Reactor (SMR) coupled to a fossil-fuel powered gas turbine.
Japan Atomic Energy Research Institute's (JAERI's) High-Temperature Test Reactor (HTTR) of 30 MWt started up at the end of 1998 and has been run successfully at 850°C for 30 days. 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 MWt 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 metres high arranged in a ring, with reflectors. Each column consists of eight one-metre 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 two-yearly refuelling, alternate layers of elements are replaced so that each remains for four years.
China's HTR-10, a 10 MWt high-temperature gas-cooled experimental reactor at the Institute of Nuclear & 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 burn-up of 80 GWday/t U. Each pebble fuel element has 5g of uranium enriched to 17% in around 8300 TRISO-coated 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.
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, the temperature never exceeded a safe 1600°C, and there was no fuel failure. This was one of six safety demonstration tests conducted then. 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.)
Construction of a larger version of the HTR-10, China's HTR-PM, was approved in principle in November 2005, with preparation for first concrete in mid 2011 and full construction start in December 2012. This was to be a single 200 MWe (450 MWt) unit but will now have twin reactors, each of 250 MWt driving a single 210 MWe steam turbine. Each reactor has a single steam generator producing steam at 566°C. The fuel is 9% enriched (520,000 elements) giving 80 GWd/t discharge burn-up. Core outlet temperature is 750ºC for the helium. 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 (see section on PBMR below). Core height is 11 metres. This 210 MWe Shidaowan demonstration plant at Rongcheng in Shandong province 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.
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$ 430 million (but later units falling to US$1500/kW with generating cost about 5 ¢/kWh). Start-up was scheduled for 2013, now 2015. 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 utilize thorium in the fuel. Eventually a series of HTRs, possibly with Brayton cycle directly driving the gas turbines, would be factory-built and widely installed throughout China.
Performance of both this and South Africa's PBMR design 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 a light water reactor, and if coolant circulation ceases the fuel will survive initial high temperatures while the reactor shuts itself down – giving inherent safety. Power control is by varying the coolant pressure, and hence flow. (See also section on Shidaowan HTR-PM in the information page on Nuclear Power in China and the Research and development section in the information page on China's Nuclear Fuel Cycle.)
South Africa's pebble bed modular reactor (PBMR) was being developed by the PBMR (Pty) Ltd consortium led by the utility Eskom, latterly with involvement of Mitsubishi Heavy Industries, and draws on German expertise. It aimed for a step change in safety, economics and proliferation resistance. Full-scale production units had been planned to be 400 MWt (165 MWe) but more recent plans were for 200 MWt (80 MWe)7. Financial constraints led to delays8 and in September 2010 the South African government confirmed it would stop funding the project9. However, a 2013 application for federal funds from National Project Management Corporation (NPMC) in the USA appears to revive the earlier direct-cycle PBMR design, emphasising its ‘deep burn’ attributes in destroying actinides and achieving high burn-up at high temperatures.
The earlier plans for the 400 MWt PBMR following a 2002 review envisaged 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 would be 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. The PBMR Demonstration Power Plant (DPP) was expected to start construction at Koeberg in 2009 and achieve criticality in 2013, but after this was delayed it was decided to focus on the 200 MWt design6
The 200 MWt (80 MWe) later design announced in 2009 was to use a conventional Rankine cycle, enabling the PBMR to deliver super-heated steam via a steam generator as well as generate electricity. This design "is aimed at steam process heat applications operating at 720°C, which provides the basis for penetrating the nuclear heat market as a viable alternative for carbon-burning, high-emission heat sources."10 An agreement with Mitsubishi Heavy Industries to take forward the R&D on this design was signed in February 2010. MHI had been involved in the project since 2001, having done the basic design and R&D of the helium-driven turbo generator system and core barrel assembly, the major components of the 400 MWt direct-cycle design.
The PBMR has a vertical steel reactor pressure vessel which contains and supports a metallic core barrel, which in turn supports the cylindrical pebble fuel core. This core is surrounded on the side by an outer graphite reflector and on top and bottom by graphite structures which provide similar upper and lower neutron reflection functions. Vertical borings in the side reflector are provided for the reactivity control elements. Some 360,000 fuel pebbles (silicon carbide-coated 9.6% enriched uranium dioxide particles encased in graphite spheres of 60 mm diameter) cycle through the reactor continuously (about six times each) until they are expended after about three years. This means that a reactor would require 12 total fuel loads in its design lifetime.
A pebble fuel plant at Pelindaba was planned. Meanwhile, the company produced some fuel which was successfully tested in Russia.
The PBMR was proposed for the US Next Generation Nuclear Plant project and submission of an application for design certification reached the pre-application review stage. The company is part of the National Project Management Corporation (NPMC) consortium which has applied for the second round of DOE funding in 2013.
PBMR development in South Africa has now been abandoned due to lack of funds. For more on it, see the PBMR Appendix in the information page on Nuclear Power in South Africa.
In the 1970s General Atomics developed an HTR with prismatic fuel blocks based on those in the 842 MWt Fort St Vrain reactor, which ran 1976-89 in the USA. Licensing review by the NRC was under way until the projects were cancelled in the late 1970s.
Evolved from this in the 1980s, General Atomics' Gas Turbine - Modular Helium Reactor (GT-MHR), would be built as modules of up to 600 MWt, but typically 350 MWt, 150 MWe. In its electrical application each would directly drive a gas turbine at 47% thermal efficiency. It can also be used for hydrogen production (100,000 t/yr claimed) and other high temperature process heat applications. The annular core, allowing passive decay heat removal, 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 the core is replaced every 18 months. Enrichment is about 15.5%, burn-up is up to 220 GWd/t, and coolant outlet temperature is 750°C with a target of 1000°C.
The GT-MHR is being developed by General Atomics in partnership with Russia's OKBM Afrikantov, supported by Fuji (Japan). Areva was formerly involved, but it has developed the basic design itself as Antares. Initially the GT-MHR was to 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 has languished since.
General Atomics says that the GT-MHR neutron spectrum is such, and the TRISO fuel is 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 would enable reactivity control and very high burn-up could be achieved with it – over 500 GWd/t – the 'Deep Burn' concept. Over 95% of the Pu-239 and 60% of other actinides would be destroyed in a single pass.
A smaller version of the GT-MHR, 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.
In February 2010, General Atomics announced a modified version of its GT-MHR design, but as a fast neutron reactor – the Energy Multiplier Module (EM2). The EM2 is a 500 MWt, 240 MWe helium-cooled fast-neutron HTR operating at 850°C and fuelled with 20 tonnes of used PWR fuel or depleted uranium, plus 22 tonnes of low-enriched uranium (~12% U-235) as starter. Used fuel from this is processed to remove fission products (about 4 tonnes) and the balance is recycled as fuel for subsequent rounds, each time topped up with 4 tonnes of further used PWR fuel. (The means of reprocessing to remove fission products is not specified.) Each refuelling cycle may be as long as 30 years. With repeated recycling the amount of original natural uranium (before use by PWR) used goes up from 0.5% to 50% at about cycle 12. High-level wastes are about 4% of those from PWR on open fuel cycle. A 48% thermal efficiency is claimed, using Brayton cycle. EM2 would also be suitable for process heat applications. The main pressure vessel can be trucked or railed to the site, and installed below ground level, and the high-speed (gas) turbine generator is also truck-transportable. The means of reprocessing to remove fission products is not specified. The company applied for the second round of DOE funding in 2013.
The company anticipates a 12-year development and licensing period, which is in line with the 80 MWt experimental technology demonstration gas-cooled fast reactor (GFR) in the Generation IV programl. GA has teamed up with Chicago Bridge & Iron, Mitsubishi Heavy Industries, and Idaho National Laboratory to develop the EM2.
Antares – Areva SC-HTGR
Another full-size HTR design is being put forward by Areva. It is based on the GT-MHR and has also involved Fuji. Reference design is 625 MWt with prismatic block fuel like the GT-MHR. Core outlet temperature is 750°C for the steam-cycle HTR version (SC-HTGR), though an eventual very high temperature reactor (VHTR) version is envisaged with 1000°C and direct cycle. The present concept uses an indirect cycle, with steam in the secondary system, or possibly a helium-nitrogen mix for VHTR, removing the possibility of contaminating the generation, chemical or hydrogen production plant with radionuclides from the reactor core. It was selected in 2012 for the US Next Generation Nuclear Plant, with 2-loop secondary steam cycle, the 625 MWt probably giving 250 MWe per unit, but the primary focus being the 750°C helium outlet temperature for industrial application.
Urenco with others commissioned a study by TU-Delft and Manchester University on the basis of which it has called for European development of very small – 5 to 10 MWe – 'plug and play' inherently-safe reactors. These are based on graphite-moderated, helium cooled HTR concepts. The fuel block design is based on that of the Fort St Vrain (FSV) reactor in USA. It would use 17-20% enriched uranium and possibly thorium fuel. A 20 MWt and a 10 MWt design have been developed, the latter with beryllium oxide reflector. The smaller "U-battery" would run for 5years before refueling and servicing, the larger one for 10 years. The smaller design, 1.8 m diameter, may be capable of being returned to the factory for this. Urenco is seeking government support for a prototype.
A small 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 is a fixed, annular bed with about 80,000 fuel elements each 6 cm 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 800°C and a thermal efficiency near 25%. Power output is controlled by limiting coolant flow. A demonstration plant is proposed for completion after 2018. The Adams Engine is deigned to be competitive with combustion gas turbines.
A small Russian HTR which was being developed by the N.A. Dollezhal Research and Development Institute of Power Engineering (NIKIET) is the modular transportable small power nuclear reactor (MTSPNR) for heat and electricity supply of remote regions. It is described as a single circuit air-cooled HTR with closed cycle gas turbine. It uses 20% enriched fuel and is designed to run for 25 years without refuelling. A twin unit plant delivers 2 MWe and/or 8 GJ/hr. No recent information is available, but an antecedent is the Pamir, from Belarus:
Since 2010 NIKIET is also involved with Luch Scientific Production Association (SPA Luch) and a Belarus organization, the Joint Institute for Power Engineering and Nuclear Research (Sosny), to design a small transportable nuclear reactor. The new design will be an HTR concept similar to Pamir but about 2.4 to 2.6 MWe.
The project draws on Sosny’s experience in designing the Pamir-630D truck-mounted small air-cooled nuclear reactor, two of which were built in Belarus from 1976 during the Soviet era. The entire plant required several trucks. This was a 600 kWe HTR reactor using 45% enriched fuel and driving a gas turbine with nitrogen tetraoxide (N2O4) through the Brayton cycle. After some operational experience the Pamir project was scrapped in 1986. It had been preceded by the 2 MWe TES-3, mounted on an extended heavy tank chassis. The prototype started up in 1961 and was abandoned in 1969.
Hybrid SMR concept
The hybrid-nuclear Small Modular Reactor (SMR) design from Hybrid Power Technologies LLC produces massive quantities of compressed air, while the gas turbine, able to burn a variety of fossil fuels, generates electrical power. Helium from the 600 MWt graphite-moderated reactor drives a primary turbine coupled to an air compressor. The very high pressure air then supercharges a combined cycle gas turbine (CCGT) driving an 850 MWe generator at 85% efficiency. The reactor and compressor are in a full containment structure. (The actual HTR is equivalent to less then 300 MWe output, so that component is still ‘small’.) The company has applied for the second round of DOE funding in 2013.
Supercritical CO2 direct cycle fast reactor concept
This is a Generation IV design based partly on the well-proven UK Advanced Gas-Cooled reactors (AGRs). The supercritical direct cycle gas fast reactor (SC-GFR) uses the supercritical CO2 coolant at 20 MPa and 650C from a fast reactor of 200 to 400 MW thermal in Brayton cycle. A small long-life reactor core could maintain decay heat removal by natural circulation. A 2011 paper from Sandia Laboratories describes it. (S-CO2 is applicable to many different heat sources, including concentrated solar. It claims high efficiency with smaller and simpler power plants. With a helium-cooled HTR or sodium-cooled fast reactor, it would be the secondary circuit.)
Fast neutron reactors
Fast neutron reactors (FNR) are smaller and simpler than light water types, they have better fuel performance and can have a longer refueling interval (up to 20 years), but a new safety case needs to be made for them, at least in the west. They are designed to use the full energy potential of uranium, rather than about one percent of it that conventional power reactors use. They 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 power regulation is achieved due to the reactivity feedback – loss of coolant flow leads to higher core temperature which slows the reaction. Fast reactors typically use boron carbide control rods.
Fuels are mostly 15-20% enriched and may be uranium nitride – UN, (U,Pu)N, (U,transuranic)N, U-Zr, or (U,Pu)Zr. Most coolant is corrosive (Pb or lead-bismuth eutectic) or flammable (Na). In the USA no enrichment plant is designed for more than 10% enrichment, but the government has 26 tonnes of HEU unallocated, and this might be blended down for fast reactors.
Small FNRs are designed to be factor-built and shipped to site on truck, train or barge and then shipped back again or to a regional fuel cycle centre at end of life. They would mostly be installed below ground level and with high surface area to volume ratio they have good passive cooling potential. Disposal is envisaged as entire units, without separate spent fuel storage, or after fuel removed for reprocessing.
A Gas-cooled Fast Reactor (GFR) concept – the EM2 – has been announced by General Atomics and is described in the HTR section above. The concept is also being pursued in the Generation IV program, with ALLEGRO being built in France.
See also Fast Neutron Reactors paper.
GE with the US national laboratories had been developing a modular liquid metal-cooled inherently-safe reactor – PRISM (Power Reactor Innovative Small Module) – under the Advanced Liquid Metal Reactor/Integral Fast Reactor (ALMR/IFR) program funded by the US Department of Energy. An antecedent was GE's fast reactor power plant for USS Seawolf 1957-58. The ALMR/IFR program was cancelled in 1994 and no US fast neutron reactor has so far been larger than 66 MWe and none has supplied electricity commercially. However, the 1994 pre-application safety evaluation report13 for the original PRISM design concluded that "no obvious impediments to licensing the PRISM design had been identified."
Today's PRISM is a GE-Hitachi (GEH) 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 (840 MWt) each, (or, earlier, three modules of 155 MWe, 471 MWt), each with one steam generator, that collectively drive one turbine generator. The pool-type modules below ground level contain the complete primary system with sodium coolant at about 500°C. The metal Pu & DU fuel is 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. The reactor is designed to use a heterogeneous metal alloy core with 42 fuel assemblies, 24 internal blanket assemblies, 33 radial blanket assemblies, 42 reflector assemblies, 48 radial shield assemblies and 6 control and shutdown assemblies. Fuel stays in the reactor about six years, with one-third removed every two years, and breeding ratio is 0.8. Used PRISM fuel is recycled after removal of fission products, though not necessarily into PRISM units.
The commercial-scale plant concept, part of an 'Advanced Recycling Center', would use three power blocks (six reactor modules) to provide 1866 MWe. An application for design certification is expected to be submitted in 2012, and a decision by GEH on building a demonstration plant was expected soon after then, though this may be brought forward if a demonstration plant is constructed at the Department of Energy's Savannah River siten. In 2011 GE Hitachi announced that it was shifting its marketing strategy to pitch the reactor directly to utilities as a way to recycle excess plutonium while producing electricity for the grid. GEH bills it as a simplified design with passive safety features and using modular construction techniques. Its reference construction schedule is 36 months.
In 2011 GE-H confirmed that it was talking with UK government agencies about the potential use of PRISM technology dispose of the UK's plutonium stockpile. GEH has launched a web portal in support of its proposal.
See also Electrometallurgical 'pyroprocessing' section in information page on Processing of Used Nuclear Fuel.
In February 2010 General Atomics announced its Energy Multiplier Module (EM2) design. EM2 is a 500 MWt, 240 MWe helium-cooled fast-neutron HTR operating at 850°C and fuelled with 20 tonnes of used PWR fuel or depleted uranium, plus 22 tonnes of low-enriched uranium (~12% U-235) as starter. It is designed to operate for 30 years without requiring refuelling. See fuller description above under HTRs.
Russia has experimented with several lead-cooled reactor designs, and has used lead-bismuth cooling for 40 years in its submarine reactors. (Pb-208 – 54% of naturally-occurring lead – is transparent to neutrons.) A significant Russian design from NIKIET is the BREST fast neutron reactor, of 700 MWt, 300 MWe, or more with lead as the primary coolant, at 540°C, supplying supercritical steam generators. The core sits in a pool of lead at near atmospheric pressure. It is inherently safe and uses a U+Pu nitride fuel. Fuel cycle is 10 months. No weapons-grade plutonium can be produced (since there is no uranium blanket), and used fuel can be recycled indefinitely, with on-site facilities. A pilot unit was planned to be built at Beloyarsk, and 1200 MWe units are planned. It is at preliminary design stage.
A smaller and newer Russian design is the Lead-Bismuth Fast Reactor (SVBR) of 280 MWt, 100 MWe, being developed by AKME-engineering and involving Gidropress in the design. It is an integral design, with 12 steam generators and 2 main circulation pumps sitting in the same Pb-Bi pool at 340-490°C as the reactor core. It is designed to be able to use a wide variety of fuels, though the pilot unit will initially use uranium oxide enriched to 16.3%. With U-Pu MOX fuel it would operate in closed cycle. Refueling interval is 7-8 years and 60-year operating life is envisaged. The melting point of the Pb-Bi coolant is 123.5°C, so it is readily kept molten during shutdown by decay heat supplemented by external heat sources if required.
The SVBR-100 unit of 280 MWt would be factory-made and transported (railway, road or waterway) as a 4.5m diameter, 8.2m high module. A power station with such modules is expected to supply electricity at lower cost than any other new technology with an equal capacity as well as achieving inherent safety and high proliferation resistance. (Russia built seven Alfa-class submarines, each powered by a compact 155 MWt Pb-Bi cooled reactor, and 80 reactor-years operational experience was acquired with these.) Overnight capital cost is previously estimated as $4000-4500 /kW and generating costs 4-5 c/kWh on 90% load factor.
In December 2009, AKME-engineering, a 50-50 joint venture, was set up by Rosatom and the En+ Group (a subsidiary of Basic Element Group) as an open joint stock company to develop and build a pilot SVBR unit14. En+ is an associate of JSC EuroSibEnergo and a 53.8% owner of Rusal, which had been in discussion with Rosatom regarding a Far East nuclear power plant and Phase II of the Balakovo nuclear plant. It was to contribute most of the capital, and Rosatom is now looking for another investor. In 2011 the EuroSibEnergo 50% share passed to its subsidiary JSC Irkutskenergo. The main project participants are OKB Gidropress at Podolsk, VNIPIET OAO at St Petersburg, and the RF State Research Centre Institute of Physics & Power Engineering (IPPE or FEI) at Obninsk. The plan is to complete the design development and put on line a 100 MWe pilot facility by 2019, with total investment of RUR16 billion ($585 million). The site is to be the Research Institute of Atomic Reactors (RIAR or NIIAR) at Dimitrovgrad – Russia's largest nuclear research centre, though earlier plans were to put it at IPPE/FEI at Obninsk. The SVBR-100 could be the first reactor cooled by heavy metal to to generate electricity. It is described by Gidropress as a multi-function reactor for power, heat or desalination.
An SVBR-10 is also envisaged, with the same design principles, a 20-year refuelling interval and generating capacity of 12 MWe, though it too is a multi-purpose unit.
(Link to SVBR brochure)
LeadCold Reactors (Blykalla Reaktorer) is a spin-off company from the Royal Institute of Technology (KTH) in Stockholm. Its SEALER (Swedish Advanced Lead Reactor) is a lead-cooled reactor designed with the smallest possible core that can achieve criticality in a fast spectrum using 20% enriched uranium oxide (UOX) fuel. The reactor is 8 MWt, with a peak electric power of 3 MWe, leading to a core life of 30 full power years (at 90% availability). The reactor vessel is designed to be small enough to permit transportation by aircraft. As the regulatory framework for licensing of small reactors in Canada is better established than in most other countries, Nunavut and the Northwest Territories are likely to become the first markets for SEALER units. The company intends to submit documentation for Phase 1 of the Canadian Nuclear Safety Commission pre-licence review in 2014.
SEALER-5 is a 5 MWe reactor design. Replacing the standard uranium oxide fuel with uranium nitride (UN), the same core can host 40% more fissile material. This allows the core to operate at 40% higher thermal power for the same duration as SEALER-3, i.e. 30 years.
SEALER-10 is the waste management system. After 30 years of operation, the early SEALER units will be transported back to a centralised recycling facility. The plutonium and minor actinides present in the spent fuel will then be separated and converted into nitride fuel for reycle in a 10 MWe SEALER reactor. One such reactor will be sufficient to manage the used fuel of ten smaller SEALER units.
Gen4 (Hyperion) Power Module
The Gen4 Module is a 70 MWt/25 MWe lead-bismuth cooled reactor concept using 19.75% enriched uranium nitride fuel, from Gen4 Energy. The reactor was originally conceived as a potassium-cooled self-regulating 'nuclear battery' fuelled by uranium hydridem. However, in 2009, Hyperion Power changed the design to uranium nitride fuel and lead-bismuth cooling to expedite design certification12. This now classes it as a fast neutron reactor, without moderation. The company claims that the ceramic nitride fuel has superior thermal and neutronic properties compared with uranium oxide. Enrichment is 19.75% and operating temperature about 500°C. The unit would be installed below ground level.
The reactor vessel housing the core and primary heat transfer circuit is about 1.5 metres wide and 2.5 metres high. It is easily portable, sealed and has no moving parts. A secondary cooling circuit transfers heat to an external steam generator. The reactor module is designed to operate for electricity or process heat (or cogeneration) continuously for up to 10 years without refuelling. Another reactor module could then take its place in the overall plant. The old module, with fuel burned down to about 15% enrichment, would be put in dry storage at site to cool for up to two years before being returned to the factory.
In March 2010, Hyperion (as the company then was) notified the US Nuclear Regulatory Commission that it planned to submit a design certification application in 2012. The company said then that it has many expressions of interest for ordering units. In September 2010, the company signed an agreement with Savannah River Nuclear Solutions to possibly build a demonstration unit at the Department of Energy site there. Hyperion planned to build a prototype by 2015, possibly with uranium oxide fuel if the nitride were not then available. In March 2012 the US DOE signed an agreement with Hyperion regarding constructing a demonstration unit at its Savannah River site in South Carolina.
In 2014 two papers on nuclear marine propulsion were published arising from a major international industry project led by Lloyd's Register. They describe a preliminary concept design study for a 155,000 dwt Suezmax tanker that is based on a conventional hull form with a 70 MW Gen4 Energy power module for propulsion.
In March 2012 Hyperion Power Generation changed its name to Gen4 Energy, and the name of its reactor to Gen4 Module (G4M). It pitched its design for remote sites having smaller power requirements.
The China Experimental Fast Reactor of 65 MWt is basically that, rather than a power reactor, though it can incidentally generate 20 MWe. It is an important part of China’s reactor development, and details are in the R&D section of the China Fuel Cycle paper. It is sodium-cooled at 530°C and has been operating since 2010.
Encapsulated Nuclear Heat-Source
The Encapsulated Nuclear Heat-Source (ENHS) is a liquid metal-cooled reactor concept of 50 MWe being developed by the University of California, Berkeley. The core is at the bottom of 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 automatic. The whole reactor sits in a 17 metre deep silo. Fuel is a uranium-zirconium alloy with 13% enrichment (or U-Pu-Zr with 11% Pu) with a 15-20 year life. After this the module is removed, stored on site until the primary lead (or Pb-Bi) 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 commercialisation.
STAR-LM, STAR-H2, SSTAR
The Secure Transportable Autonomous Reactor (STAR) project at Argonne National Laboratory is developing small, multi-purpose systems that operate nearly autonomously for the very long term. The STAR-LM is a factory-fabricated fast neutron modular reactor cooled by lead-bismuth eutectic, with passive safety features. Its 300-400 MWt size means it can be shipped by rail. It uses uranium-transuranic nitride fuel in a 2.5 m diameter cartridge which is replaced every 15 years. Decay heat removal is by external air circulation. The STAR-LM was conceived for power generation with a capacity of about 175 MWe.
The STAR-H2 is an adaptation of the same reactor 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). Its development is further off.
A smaller STAR variant is the Small Sealed Transportable Autonomous Reactor (SSTAR) which was being developed by Lawrence Livermore, Argonne and Los Alamos National Laboratories in collaboration with others including Toshiba. It has lead or Pb-Bi cooling, 564°C core outlet temperature and has integral steam generator inside the sealed unit, which would be installed below ground level. Conceived in sizes 10-100 MWe, main development was focused on a 45 MWt/20 MWe version as part of the US Generation IV effort. After a 20 or 30-year life without refuelling, the whole reactor unit is then returned for recycling the fuel. The reactor vessel is 12 metre high and 3.2 m diameter and the core one metre high and 1.2 m diameter (20 MWe version). SSTAR would eventually be coupled to a Brayton cycle turbine using supercritical carbon dioxide with natural circulation to four heat exchangers. A prototype was envisaged for 2015, but development has apparently ceased.
Integral Fast Reactor, ARC-100
Advanced Reactor Concepts LLC (ARC) is commercializing a 100 MWe sodium-cooled fast reactor based on the 62.5 MWt Experimental Breeder Reactor II (EBR-II). The EBR-II was significant fast reactor prototype at Idaho National Laboratory (formerly Argonne National Laboratory – West) which produced 19 MWe over about 30 years. It used the pyrometallurgically-refined used fuel from light water reactors as fuel, including a wide range of actinides. After operating 1963 to 1994 it is being decommissioned. EBR-II was the basis of the US Integral Fast Reactor (IFR) program (originally the Advanced Liquid Metal Reactor program), and that term is again in use. An EBR-III of 200-300 MWe was proposed but not developed (see also information page on Fast Neutron Reactors).
The ARC-100 system comprises a uranium alloy core submerged in sodium. The liquid sodium is passed through the core where it is heated to 510°C, then passed through an integral heat exchanger (within the pool) where it heats sodium in an intermediate loop, which in turn heats working fluid for electricity generation. It would have a refueling interval of 20 years. A 50 MWe version of the ARC is also under development.
A lead-bismuth-eutectic (LBE) cooled fast reactor of 150 MWt /53 MWe, the LSPR (LBE-Cooled Long-Life Safe Simple Small Portable Proliferation-Resistant Reactor), is under development in Japan. Fuelled units would be supplied from a factory and operate for 30 years, then be returned. The concept is intended for developing countries.
A small-scale design developed by Japan's Central Research Institute of Electric Power Industry (CRIEPI) in cooperation with Mitsubishi Research Institute and funded by the Japan Atomic Energy Research Institute (JAERI) is the 5 MWt, 200 kWe Rapid-L, using lithium-6 (a neutron poison) as control medium. It would have 2700 fuel pins of 40-50% enriched uranium nitride with 2600°C melting point integrated into a disposable cartridge or 'integrated fuel assembly'. The reactivity control system is passive, using lithium expansion modules (LEMs) which give burn-up compensation, partial load operation as well as negative reactivity feedback. During normal operation, lithium-6 in the LEM is suspended on an inert gas above the core region. As the reactor temperature rises, the lithium-6 expands, moving the gas/liquid interface down into the core and hence adding negative reactivity. Other kinds of lithium modules, also integrated into the fuel cartridge, shut down and start up the 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 metres high and 2 metres diameter.
The larger RAPID reactor delivers 1 MWe and is U-Pu-Zr fuelled and sodium-cooled.
The Super-Safe, Small & Simple (4S) 'nuclear battery' system is being developed by Toshiba and the Central Research Institute of Electric Power Industry (CRIEPI) in Japan in collaboration with SSTAR work and Westinghouse (owned by Toshiba) in the USA. It uses sodium as coolant (with electromagnetic pumps) and has passive safety features, notably negative temperature coefficient of 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 30 MWt (10 MWe) version or 11.5% Pu for the 135 MWt (50 MWe) version. Steady power output over the core lifetime in 30 MWt version is achieved by progressively moving upwards an annular reflector around the slender core (0.68m diameter, 2m high in the small version; 1.2m diameter and 2.5m high in the larger version) at about one millimetre per week. 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. Burn-up will be 34 GWday/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 versions of 4S are designed to automatically maintain an outlet coolant temperature of 510-550ºC – 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 10 MWe (30 MWt) 4S reactor in that remote location. A pre-application Nuclear Regulatory Commission (NRC) review is under way with a view to application for design certification in October 2010 (delayed from 2009 by NRC workload), and combined construction and operating licence (COL) application to follow. Its design is sufficiently similar to PRISM – GE's modular 150 MWe liquid metal-cooled inherently-safe reactor which went part-way through the NRC approval process (see section below on PRISM) – for it to have good prospects of licensing. Toshiba plans a worldwide marketing program to sell the units for power generation at remote mines, for extraction of tar sands, desalination plants and for making hydrogen. Eventually it expects sales for hydrogen production to outnumber those for power supply.
The L-4S is a Pb-Bi cooled version of the 4S design.
Travelling wave and standing wave reactors
Following work by the DOE's Lawrence Livermore Laboratory, a 1950s design concept resurfaced in 1996 as the travelling wave reactor (TWR). This has been considered in the past as, generically, a candle reactor, or breed-burn reactor, since it is designed to burn slowly from one end of a core to the other, making the actual fuel as it goes. Having started with a small amount of enriched uranium, the reactor would run on natural or depleted uranium packed inside hundreds of hexagonal pillars. In a 'wave' that moved through the core at only one centimetre per year, the U-238 would be bred progressively into Pu-239, which is the actual fuel and undergoes fission. In 2009 this was boldly selected by MIT Technology Review as one of ten emerging technologies of note15. Eventual sizes could range from a few hundred MWe to over 1000 MWe. Microsoft founder Bill Gates is providing financial backing for Terrapower, a company founded in 2008.
However, by mid-2011 Terrapower changed the design to be a standing wave reactor, since too many neutrons would be lost behind the travelling wave of the previous design and it would not be practical to remove the heat efficiently – the cooling system could not follow the wave. A standing wave design would start the fission reaction at the centre of the reactor core, where the breeding stays, and operators would move fresh fuel assemblies from the outer edge of the core progressively to the central region to catch neutrons, while shuffling spent fuel out of the centre to the periphery. It is thus more like a standard sodium-cooled fast reactor, albeit without fresh fuel being added, and could reach a fuel burn-up of "up to 30%". Fuel is metal, and the coolant is sodium. Terrapower says a 600 MWe demonstration plant – TWR-P – is planned for 2018-22 construction followed by operation of larger commercial plants of about 1150 MWe from late 2020s.
Further details in Fast Neutron Reactors paper.
Korean fast reactor designs
In South Korea, the Korea Atomic Energy Research Institute (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 design 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)
These use molten fluoride salts as primary coolant, at low pressure. Lithium-beryllium fluoride and lithium fluoride salts remain liquid without pressurization up to 1400°C, in marked contrast to a PWR which operates at about 315°C under 150 atmospheres pressure. In most designs (not the AHTR) the fuel is dissolved in the coolant.
During the 1960s, the USA developed the molten salt breeder reactor concept as the primary back-up option for the fast breeder reactor (cooled by liquid metal) and a small prototype 8 MWt Molten Salt Reactor Experiment (MSRE) operated at Oak Ridge over four years to 1969 (the MSR program ran 1957-1976). U-235 tetrafluoride enriched to 33% was in molten lithium, beryllium and zirconium fluorides at 860°C which flowed through a graphite moderator. A second campaign used U-233 fuel, but the program did not progress to building a MSR breeder utilising thorium. There is now renewed interest in the concept in Japan, Russia, China, France and the USA, and one of the six Generation IV designs selected for further development is the molten salt reactor (MSR).
In the normal MSR, the fuel is a molten mixture of lithium and beryllium fluoride (FLiBe) salts with dissolved enriched uranium (U-235) or U-233 fluorides (UF4). The core consists of unclad graphite moderator arranged to allow the flow of salt at some 700°C and at low pressure. Much higher temperatures are possible but not yet tested. Heat is transferred to a secondary salt circuit and thence to steamo. The basic design is not a fast neutron reactor, but with some moderation by the graphite is epithermal (intermediate neutron speed) and breeding ratio is less than 1. Thorium can be dissolved with the uranium in a single fluid MSR, known as a homogeneous design. Two-fluid, or heterogeneous MSRs would have fertile salt containing thorium in a second loop separate from the fissile salt containing uranium and could operate as a breeder reactor (MSBR).
The fission products dissolve in the salt and are removed continuously in an on-line reprocessing loop and replaced with fissile uranium or, potentially, Th-232 or U-238. Actinides remain in the reactor until they fission or are converted to higher actinides which do so.
Lithium used in the salt must be fairly pure Li-7, since Li-6 produces tritium when fissioned by neutrons. Li-7 has a very small neutron cross section. This means that natural lithium must be enriched.
The concept is being pursued in the Generation IV program with two variants: one a fast neutron reactor with fissile material dissolved in the circulation fuel salt, and with solid particle fuel in graphite and the salt functioning only as coolant.
See also Molten Salt Reactor paper.
Liquid Fluoride Thorium Reactor
The Liquid Fluoride Thorium Reactor (LFTR) is a heterogeneous MSR design which breeds its U-233 fuel from a fertile blanket of of lithium-beryllium fluoride salts with thorium fluoride. Some of the neutrons released during fission of the U-233 salt in the reactor core are absorbed by the thorium in the blanket salt. The resulting U-233 is separated from the blanket salt and in lithium and beryllium fluorides becomes the liquid core fuel. LFTRs can rapidly change their power output, and hence be used for load-following.
Because they are expected to be inexpensive to build and operate, 100 MWe LFTRs could be used as peak and back-up reserve power units. They would normally operate at much higher temperatures than LWRs – up to at least 700°C, and hence have potential for process heat. China is building a 5 MWe thorium-breeding molten-salt reactor (Th-MSR or TMSR), essentially an LFTR, with 2015 target for operation at the Shanghai Institute of Nuclear Applied Physics (SINAP). China claims to have the world's largest national effort on these and hopes to obtain full intellectual property rights on the technology. The US Department of Energy is collaborating with the China Academy of Sciences on the program, which had a start-up budget of $350 million. The target date for TMSR deployment is 2032.
Molten fluoride salts (possibly simply cryolite – Na-Al fluoride) are a preferred interface fluid in a secondary circuit between the nuclear heat source and any chemical plant. The aluminium smelting industry provides substantial experience in managing them safely. The hot molten salt can also be used with secondary helium coolant generating power via the Brayton cycle.
The Fuji MSR is a 100-200 MWe design to operate as a near-breeder and being developed internationally by a Japanese, Russian and US consortium: the International Thorium Molten Salt Institute (ITHMSI). 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 (Pu-242 being the dominant Pu isotope); low fuel use (the French self-breeding variant claims 50kg of thorium and 50kg U-238 per billion kWh); and safety due to passive cooling up to any size. A 10 MWe mini Fuji is also on the drawing board.
AHTR / FHR
Research on molten salt coolant has been revived at Oak Ridge National Laboratory ORNL) in the USA with the Advanced High-Temperature Reactor (AHTR).16 This is a larger reactor using a coated-particle graphite-matrix fuel like that in the GT-MHR (see above section on the GT-MHR) and with molten fluoride salt as primary coolant. It is also known as the Fluoride High Temperature Reactor (FHR). 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 FLiBe salt is used solely as coolant, and achieves temperatures of 750-1000°C or more while at low pressure. This could be used in thermochemical hydrogen manufacture.
A 5 MW thorium-fueled prototype is under construction at Shanghai Institute of Nuclear Applied Physics (SINAP, under the China Academy of Sciences) with 2015 target for operation. A 100 MWt demonstration pebble-bed plant with open fuel cycle is planned by about 2025. SINAP sees this design having potential for higher temperatures than MSRs with fuel salt.
In the USA a consortium including UC Berkeley, ORNL and Westinghouse is designing a 100 MWe pebble-bed FHR, with modular construction and able to deliver 240 MWe with gas co-firing. A 410 MWe/900 MWt pebble bed version was also being designed with UC-Berkeley.
Reactor sizes of 1500 MWe/3600 MWt are envisaged, with capital costs estimated at less than $1000/kW.
In the secondary cooling circuit air is compressed, heated, flows through gas turbines producing electricity, enters a steam recovery boiler producing steam that produces additional electricity, and exits to the atmosphere. Added peak power can be produced by injecting natural gas (or hydrogen in the future) after nuclear heating of the compressed air to raise gas temperatures and plant output, giving it rapidly variable output (of great value in grid stability and for peak load demand where renewables have significant input). This is described as an air Brayton combined-cycle (ABCC) system in secondary circuit.
Canada-based Terrestrial Energy Inc (TEI) has designed the Integral MSR. This simplified MSR integrates the primary reactor components, including primary heat exchangers to secondary clean salt circuit, in a sealed and replaceable core vessel that has a projected life of seven years. The IMSR will operate at approximately 700°C, which can support many industrial process heat applications. The moderator is a hexagonal arrangement of graphite elements. The fuel-salt is a eutectic of low-enriched uranium fuel (UF4) and a fluoride carrier salt at atmospheric pressure. Emergency cooling and residual heat removal are passive. The IMSR is designed in three sizes: 80 MWth, 300 MWth, and 600 MWth. The total levelized cost of electricity from the largest is projected to be competitive with natural gas. The smallest is designed for off-grid, remote power applications. The company hopes to commission its first commercial reactor by the early 2020s.
Transatomic Power Corp is a new US company partly funded by Founders Fund and aiming to develop a MSR using very low-enriched uranium fuel (1.8%) or the entire actinide component of used LWR fuel. The TAP reactor has an efficient zirconium hydride* moderator and a LiF-based fuel salt bearing the UF4, hence a very compact core. Owing to the ZrH moderator, there are significantly more neutrons in the thermal region (less than 1 eV) compared with a graphite moderator, thereby enabling the reactor to generate power from very low-enriched uranium or used LWR fuel. The epithermal (1 eV - 1 MeV) spectrum is lower than that with graphite, but in the fast spectrum (over 1 MeV) the neutron flux is greater than with graphite moderator, and therefore contributes strongly to actinide burning.
The envisaged first commercial plant will be 1250 MWt/550 MWe running at 44% thermal efficiency with 650°C in primary loop, using steam cycle via an intermediate loop with FLiNaK salt (LiF-KF-NaF). It would give up to 96% actinide burn-up. It has negative void and thermal coefficients, and the moderator starts to fail at higher temperatures due to hydrogen loss. Decay heat removal can be by convection. The overnight cost for an nth-of-a-kind 550 MWe plant, including lithium-7 inventory and on-line fission product removal and storage, is estimated at $2 billion with a three-year construction schedule. A version of the reactor may utilize thorium fuel.
* as used in TRIGA research reactors and TOPAZ and SNAP reactors for space program.
Flibe Energy has a 40 MW thermal reactor design that uses lithium fluoride/beryllium fluoride (FLiBe) salt as its coolant. This is based on earlier US work on the Molten-Salt Reactor Program.
Aqueous homogeneous reactors
Aqueous homogeneous reactors (AHRs) have the fuel mixed with the moderator as a liquid. Typically, low-enriched uranium nitrate is in aqueous solution. About 30 AHRs have been built as research reactors and have the advantage of being self-regulating and having the fission products continuously removed from the circulating fuel. A 1 MWt AHR operated in the Netherlands 1974-77 using Th-HEU MOX fuel. Further detail is in the Research Reactors paper.
A theoretical exercise published in 2006 showed that the smallest possible thermal fission reactor would be a spherical aqueous homogenous one powered by a solution of Am-242m(NO3)3 in water. Its mass would be 4.95 kg, with 0.7 kg of Am-242m nuclear fuel, and diameter 19 cm. Power output would be a few kilowatts. Possible applications are space program and portable high-intensity neutron source. The small size would make it easily shielded.
This is a new design from Northern Nuclear Industries in Canada, combining a number of features in unique combination. The 100 MWt, 36 MWe reactor has a graphite moderator, TRISO fuel in pebbles, lead (Pb-208) as primary coolant, all as integral pool-type arrangement at near atmospheric pressure. It delivers steam at 370°C, and is also envisaged as an industrial heat plant. The fuel pebbles are in four cells, each with graphite reflectors, and capacity can be increased by adding cells. Shutdown rods are similar to those in CANDU reactors. Passive decay heat removal is by air convection. The company present it as a Gen IV design
Modular construction using small reactor units
Westinghouse and IRIS partners have outlined the economic case for modular construction of their IRIS design (about 330 MWe), and the argument applies similarly to other similar or smaller units. They pointed out that IRIS with its size and simple design is ideally suited for modular construction in the sense of progressively building a large power plant with multiple small operating units. The economy of scale is replaced here with the economy of serial production of many small and simple components and prefabricated sections. They expected that construction of the first IRIS unit would 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, for instance, three IRIS single modules providing 1000 MWe capacity would be similar or smaller in size than one with a comparable total power single unit.
Many small reactors are designed with a view to serial construction and collective operation as modules of a large plant. In this sense they are 'small modular reactors' – SMRs – but not all small reactors are of this kind (e.g. the Toshiba 4S), though the term SMR tends to be used loosely for all small designs.
Eventually plants comprising a number of SMRs are 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 estimated 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.
a. In USA, UK, France, Russia, China, and India, mostly using high-enriched fuel. Reactors built as neutron sources are not designed to produce heat or steam, and are less relevant here. [Back]
b. A very general rule is that no single unit should be larger than 15% of grid capacity [Back]
c. 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. Because small reactors have a higher surface area to volume (and core heat) ratio compared with large units, a lot of the engineering for safety (including heat removal in large reactors) is not needed in the small ones. [Back]
d. In 2010, the American Nuclear Society convened a special committee to look at licensing issues with SMRs in the USA, where dozens of land-based small reactors were built since the 1950s through to the 1980s, proving the safety and security of light water-cooled, gas‐cooled, and metal‐cooled SMR technologies. The committee had considerable involvement from SMR proponents, along with the Nuclear Regulatory Commission, Department of Energy laboratories and universities – a total of nearly 50 individuals. The committee's interim report1 includes the following two tables, which highlight some of the differences between the established US reactor fleet and SMRs.
Comparison of current-generation plant safety systems to potential SMR design
|Current‐generation safety‐related systems
||SMR safety systems
|High‐pressure injection system.
Low‐pressure injection system.
|No active safety injection system required. Core cooling is maintained using passive systems.
|Emergency sump and associated net positive suction head (NPSH) requirements for safety‐related pumps.
||No safety‐related pumps for accident mitigation; therefore, no need for sumps and protection of their suction supply.
|Emergency diesel generators.
||Passive design does not require emergency alternating‐current (AC) power to maintain core cooling. Core heat removed by heat transfer through vessel.
|Active containment heat systems.
||None required because of passive heat rejection out of containment.
|Containment spray system.
||Spray systems are not required to reduce steam pressure or to remove radioiodine from containment.
|Emergency core cooling system (ECCS) initiation, instrumentation and control (I&C) systems. Complex systems require significant amount of online testing that contributes to plant unreliability and challenges of safety systems with inadvertent initiations.
||Simpler and/or passive safety systems require less testing and are not as prone to inadvertent initiation.
|Emergency feedwater system, condensate storage tanks, and associated emergency cooling water supplies.
||Ability to remove core heat without an emergency feedwater system is a significant safety enhancement.
Comparison of current-generation plant support systems to potential SMR design
|Current LWR support systems
||SMR support systems
|Reactor coolant pump seals. Leakage of seals has been a safety concern. Seal maintenance and replacement are costly and time‐consuming.
||Integral designs eliminate the need for seals.
|Ultimate heat sink and associated interfacing systems. River and seawater systems are active systems, subject to loss of function from such causes as extreme weather conditions and bio‐fouling.
||SMR designs are passive and reject heat by conduction and convection. Heat rejection to an external water heat sink is not required.
|Closed cooling water systems are required to support safety‐ related systems for heat removal of core and equipment heat.
||No closed cooling water systems are required for safety‐related systems.
|Heating, ventilating, and air‐conditioning (HVAC). Required to function to support proper operation of safety‐related systems.
||The plant design minimizes or eliminates the need for safety‐related room cooling eliminating both the HVAC system and associated closed water cooling systems.
Some of the early (1950s-1980) small power reactors were developed so as to provide an autonomous power source (ie not requiring continual fuel delivery) in remote areas. The USA produced eight such experimental reactors 0.3 to 3 MWe, deployed in Alaska, Greenland and Antarctica. The USSR produced about 20, of many kinds, and one (Gamma) still operates at the Kurchatov Institute. Another is the Belarus Pamir, mentioned in the HTR section above. [Back]
e. The first two-unit VBER-300 plant was planned to be built in Aktau city, western Kazakhstan, with completion of the first unit originally envisaged in 2016, and 2017 for the second. The Kazakhstan-Russian Nuclear Stations joint stock company (JSC) was established by Kazatomprom and Atomstroyexport (on a 50:50 basis) in October 2006 for the design, construction and international marketing of the VBER-300. See page on the VBER-300 on the Kazatomprom website (www.kazatomprom.kz) [Back]
f. The 200 MWt (50 MWe net) Melekess VK-50 prototype BWR in Dimitrovgrad, Ulyanovsk commenced operation in 1965 [Back]
g. Central Argentina de Elementos Modulares (CAREM). See the Invap website (www.invap.com.ar) [Back]
h. The page on the NHR-5 on the website of Tsingua University's Institute of Nuclear Energy Technology (now the Institute of Nuclear and New Energy Technology, www.inet.tsinghua.edu.cn) describes the NHR-5 as "a vessel type light water reactor with advanced features, including integral arrangement, natural circulation, hydraulic control rod driving and passive safety systems. Many experiments have been conducted on the NHR-5, such as heat-electricity cogeneration, air-conditioning and seawater desalination." [Back]
i. See the page on Modular Nuclear Reactors on the Babcock & Wilcox website (www.babcock.com) [Back]
j. The 69 fuel assemblies are identical to normal PWR ones, but at about 1.7 m long, a bit less than half the length. [Back]
k. Between 1966 and 1988, the AVR (Arbeitsgemeinschaft VersuchsReaktor) experimental pebble bed reactor at Jülich, Germany, operated for over 750 weeks at 15 MWe, most of the time with thorium-based fuel (mixed with high-enriched uranium). The fuel consisted of about 100,000 billiard ball-sized fuel elements. Maximum burn-ups of 150 GWd/t were achieved. It was used to demonstrate the inherent safety of the design due to negative temperature coefficient: reactor power fell rapidly when helium coolant flow was cut off.
The 300 MWe THTR (Thorium HochTemperatur Reaktor) in Germany was developed from the AVR and operated between 1983 and 1989 with 674,000 pebbles, over half containing Th/HEU fuel (the rest graphite moderator and some neutron absorbers). These were continuously recycled and on average the fuel passed six times through the core. Fuel fabrication was on an industrial scale. The reactor was shut down for sociopolitical reasons, not because of technical difficulties, and the basic concept with inherent safety features of HTRs was again proven. It drove a steam turbine.
The 200 MWt (72 MWe) HTR-modul was then designed by Siemens/Interatom as a modular unit to be constructed in pairs, with a core height three times its diameter, allowing passive cooling for removal of decay heat, eliminating the need for emergency core cooling systems. It was licensed in 1989, but was not constructed. This design was part of the technology bought by Eskom in 1996 and is a direct antecedent of the pebble bed modular reactor (PBMR).
During 1970s and 1980s Nukem manufactured more than 250,000 fuel elements for the AVR and more than one million for the THTR. In 2007, Nukem reported that it had recovered the expertise for this and was making it available as industry support.
In addition to these pebble bed designs, the 20 MWt Dragon reactor ran in UK 1964-75, the 115 MWt Peach Bottom reactor in USA ran 1966-74, and 8432 MWt Fort St Vrain ran 1976-89 - all with prismatic fuel, and the last two supplying power commercially. In the USA the Modular High-Temperature Gas-cooled reactor (MHTGR) design was developed by General Atomics in the 1980s, with inherent safety features, but the DOE project ended in 1993. [Back]
l. The 80 MWt ALLEGRO demonstration GFR is planned by Euratom to incorporate all the architecture and the main materials and components foreseen for the full-sized GFR but without the direct (Brayton) cycle power conversion system. It is being developed mainly by France, with Japan and Switzerland, and operation about 2020 is envisaged. [Back]
m. The Hyperion Power Module was originally designed by Los Alamos National Laboratory as a 70 MWt 'nuclear battery' that uses uranium hydride (UH3) fuel, which also functions as a moderator. UH3 stores vast quantities of hydrogen, but this stored hydrogen dissociates as the temperature rises above the operating temperature of 550°C. The release of hydrogen gas lowers the density of the UH3, which in turn decreases reactivity. This process is reversed as the core temperature drops, leading to the reabsorption of hydrogen. The consequent increase in moderator density results in an increase in core reactivity11. All this is without much temperature change since the main energy gain or loss is involved in phase change. [Back]
n. In October 2010, GEH announced it was exploring the possibility with Savannah River Nuclear Solutions of building a prototype PRISM reactor at the Department of Energy’s Savannah River Site. [Back]
o. As MSRs will normally operate at much higher temperatures than LWRs, they have potential for process heat. Another option is to have a secondary helium coolant in order to generate power via the Brayton cycle. [Back]
p. Most Air Cooled Condenser (ACC) technology has a limitation in that the tubes carrying the steam must be made of carbon steel which severely limits the service life of the ACC. Holtec has developed an ACC with stainless steel tubes bonded to aluminum fins and thus with much longer service life. [Back]
1. Interim Report of the American Nuclear Society President's Special Committee on Small and Medium Sized Reactor (SMR) Licensing Issues, American Nuclear Society (July 2010) [Back]
2. Reactors ready for floating plant, World Nuclear News (7 August 2009) [Back]
3. B&W introduces scalable, practical nuclear energy, Babcock & Wilcox press release (10 June 2009); Small Reactors Generate Big Hopes, Wall Street Journal (18 February 2010) [Back]
4. Russia plans deployment of small reactors, World Nuclear News (13 September 2007) [Back]
6. Tennessee Valley Authority (TVA) – Key Assumptions Letter for the Possible Launching and Construction of Small Modular Reactor Modules at the Clinch River Site, TVA letter to the Nuclear Regulatory Commission (5 November 2010) [Back]
7. PBMR Considering Change In Product Strategy, PBMR (Pty) news release (5 February 2009) [Back]
8. PBMR postponed, World Nuclear News (11 September 2009) [Back]
9. Address by the Minister of Public Enterprises, Barbara Hogan, to the National Assembly, on the Pebble Bed Modular Reactor, Department of Public Enterprises press release (16 September 2010) [Back]
10. South Africa’s Pebble Bed Company Joins Forces with MHI of Japan, PBMR (Pty) news release (4 February 2010) [Back]
11. High hopes for hydride, Nuclear Engineering International (January 2009) [Back]
12. Hyperion launches U2N3-fuelled, Pb-Bi-cooled fast reactor, Nuclear Engineering International (November 2009) [Back]
13. Preapplication Safety Evaluation Report for the Power Reactor Innovative Small Module (PRISM) Liquid-Metal Reactor – Final Report, NUREG-1368, Office of Nuclear Reactor Regulation, US Nuclear Regulatory Commission (February 1994) [Back]
14. Initiative for small fast reactors, World Nuclear News (4 January 2010); En+ Group and Rosatom Form JV To Create Fast Neutron Reactor, En+ Group press release (25 December 2009) [Back]
15. TR10: Traveling-Wave Reactor, Matthew L. Wald, MIT Technology Review (March/April 2009); Special Report: 10 Emerging Technologies 2009, MIT Technology Review [Back]
16. The Advanced High-Temperature Reactor: High-Temperature Fuel, Molten Salt Coolant, and Liquid-Metal-Reactor Plant, Charles Forsberg, Oak Ridge National Laboratory, presented at the 1st International Conference on Innovative Nuclear Energy Systems for Sustainable Development of the World (COE INES-1) held at the Tokyo Institute of Technology, Tokyo, Japan (31 October - 4 November 2004) [Back]
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The Need for Innovative Nuclear Reactor and Fuel Cycle Systems, Victor Mourogov, presented at the 25th Annual International Symposium 2000 of The Uranium Institute, London (31 August - 1 September 2000)
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Nuclear Seawater Desalination Plant Coupled with 200 MW Heating Reactor, Haijun Jia and Yajun Zhang, Institute of Nuclear Energy Technology (INET), Tsinghua University, Beijing, China, presented at the International Symposium on the Peaceful Applications of Nuclear Technology in the Gulf Co-operation Council (GCC) Countries, Jeddah, Saudi Arabia (3-5 November 2008)
Floating Power Sources Based on Nuclear Reactor Plants, Panov et al., Federal State Unitary Enterprise the Federal Scientific and Industrial Center I. I. Afrikantov Experimental Design Bureau of Mechanical Engineering, Nizhny Novgorod, Russia, presented at the 5th International Conference on Asian Energy Cooperation: Mechanisms, Risks, Barriers (AEC-2006), organized by the Energy Systems Institute of the Russian Academy of Sciences and held in Yakutsk, Russia (27-29 June 2006)
Nuclear Desalination Complex with VK-300 Boiling-Type Reactor Facility, B.A. Gabaraev, Yu.N. Kuznetzov, A.A. Romenkov and Yu.A. Mishanina, presented at the 2004 World Nuclear Association Annual Symposium, London (8-10 September 2004)
Section on Flexblue on the DCNS website (www.dcnsgroup.com)
NuScale Power website (www.nuscalepower.com)
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High-temperature gas-cooled reactors
HTTR Home Page page on the Japan Atomic Energy Agency website (www.jaea.go.jp)
PBMR website (www.pbmr.com)
Pebble Bed Modular Reactor – The First Generation IV Reactor To Be Constructed, Sue Ion, David Nicholls, Regis Matzie and Dieter Matzner, presented at the 2003 World Nuclear Association Annual Symposium, London (3-5 September 2003)
Status of the GT-MHR for Electricity Production, M. P. LaBar, A. S. Shenoy, W. A. Simon and E. M. Campbell, presented at the 2003 World Nuclear Association Annual Symposium, London (3-5 September 2003)
GT-MHR page on the General Atomics Energy Group website (www.ga.com/energy)
EM2 page on the General Atomics Energy Group website (www.ga.com/energy)
High and very high temperature reactors page on the Areva website (www.areva.com)
Adams Atomic Engines, Inc. website (www.atomicengines.com)
HTGR Advances in China, Xu Yuanhui, Nuclear Engineering International (March 2005)
High Temperature Gas-Cooled Reactors: Lessons Learned Applicable to the Next Generation Nuclear Plant, Beck J.M. & Pinnock L.F. Idaho National Laboratory, April 2011.
Liquid metal-cooled fast reactors
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STAR-H2: Secure Transportable Autonomous Reactor for Hydrogen Production & Desalinization, Wade et al., presented at the Tenth International Conference on Nuclear Engineering (ICONE 10) held in Arlington, Virginia USA, (14-18 April 2002)
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Molten salt reactors, AHTR
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Aqueous homogeneous reactors
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Some of the developments described in this paper are fascinating and exciting. Nevertheless it is salutary to keep in mind the words of the main US pioneer in nuclear reactor development. Admiral Hyman Rickover in 1953 - about the time his first test reactor in USA started up - made some comments about "academic paper-reactors" vs. real reactors. See: http://en.wikiquote.org/wiki/Hyman_G._Rickover for the full quote:
"An academic reactor or reactor plant almost always has the following basic characteristics: (1) It is simple. (2) It is small. (3) It is cheap. (4) It is light. (5) It can be built very quickly. (6) It is very flexible in purpose. (7) Very little development will be required. It will use off-the-shelf components. (8) The reactor is in the study phase. It is not being built now.
"On the other hand a practical reactor can be distinguished by the following characteristics: (1) It is being built now. (2) It is behind schedule. (3) It requires an immense amount of development on apparently trivial items. (4) It is very expensive. (5) It takes a long time to build because of its engineering development problems. (6) It is large. (7) It is heavy. (8) It is complicated.
"The tools of the academic designer are a piece of paper and a pencil with an eraser. If a mistake is made, it can always be erased and changed. If the practical-reactor designer errs, he wears the mistake around his neck; it cannot be erased. Everyone sees it. The academic-reactor designer is a dilettante. ......."
USS Nautilus was launched in 1955.