- This chapter was published on “Inuitech – Intuitech Technologies for Sustainability” on January 18, 2011; and
- It was updated on 6 January 2016.
The International Atomic Energy Agency (IAEA) defines “small” as under 300 MWe, but according to the World Nuclear Association, in general, today 500 MWe might be, considered an upper limit to “Small Nuclear Power Reactors”.
As nuclear power generation has become, established, since the 1950s, the size of reactor units has grown from 60 MWe to more than 1600 MWe, with corresponding economies of scale in operation. At the same time there have been many hundreds of smaller reactors built both for naval use (up to 190 MW thermal) and as neutron sources, yielding enormous expertise in the engineering of deliberately small units.
In addition to the main interest in small and simpler reactors for generating clean, safe, and affordable electricity from nuclear power and for process heat is, driven both by a desire to reduce capital costs as well as to minimize the dependence upon large grid systems, here are some other reasons:
- The flexibility as small reactors may be, built independently or as modules in a large complex with the option to increase capacity incrementally as required;
- Small reactors are best suited for remote sites;
- Small reactors for power generation offer greater simplicity of design, economy of mass production, and reduced sitting costs; and
- Small reactors are designed for a high level of passive or inherent safety in the event of malfunction.
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.
Here are the highlights of the activities on small nuclear power reactors:
- The most advanced modular project is in China, where Chinergy is preparing to build the 210 MWe HTR-PM, which consists of twin 250 MWt reactors;
- In South Africa, Pebble Bed Modular Reactor (Pty) Limited and Eskom had been developing the pebble bed modular reactor (PBMR) of 200 MWt (80 MWe), but funding for this project has been stopped;
- A US group led by General Atomics is developing another design – the gas turbine modular helium reactor (GT-MHR) – with 600 MWt (285 MWe) modules driving a gas turbine directly, using helium as a coolant and operating at very high temperatures;
- 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;
- 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; and
- Also in the small reactor category are the Indian 220 MWe pressurized heavy water reactors (PHWR) 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.
Small nuclear power reactors are, classified into the following six categories as per the World Nuclear Association (Updated 12 October 2010):
US experience of small light water reactors (LWR) 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 78 million kWh. There was also an Army program for small reactor development, and some successful small reactors from the main national program commenced in the 1950s. One was the Big Rock Point BWR of 67 MWe, which operated for 35 years to 1997.
The following designs, the KLT and VBER have conventional pressure vessels plus external steam generators (PV/loop design):
Russia’s KLT-40S from OKBM Afrikantov is a reactor well proven in icebreakers and now proposed for wider use in desalination and, on barges, for remote area power supply. Here a 150 MWt unit produces 35 MWe (gross) as well as up to 35 MW of heat for desalination or district heating (or 38.5 MWe gross if power only). These are, designed to run 3-4 years between refueling with on-board refueling 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 percent capacity factor). Although the reactor core is normally, cooled by forced circulation, the design relies on convection for emergency cooling. Fuel is uranium aluminum silicide with enrichment levels of up to 20 percent, giving up to four-year refueling intervals.
The first floating nuclear power plant, the Akademik Lomonosov, commenced construction in 2007 and is, planned to be located near to Vilyuchinsk. The plant is due to be, completed in 2011;
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 percent has burnable poison; it has low burn-up (31 GWd/t average, 41.6 GWd/t maximum) and eight-year refueling interval; and
OKBM Afrikantov’s larger VBER-300 PWR is a 295 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 percent capacity factor. It has four steam generators and a cassette core with 85 fuel assemblies enriched to 5 percent and 48 GWd/tU burn-up. Versions with three and two steam generators are, also envisaged, of 230 and 150 MWe respectively. In addition, with more sophisticated and higher-enriched (18 percent) fuel in the core, the refueling interval can be pushed from two years out to 15 years with burn-up to 125 GWd/tU. A 2006 joint venture between Atomstroyexport and Kazatomprom sets this up for development as a basic power source in Kazakhstan, then for export.
The following thirteen nuclear power reactor designs under this category, have, the steam supply system inside the reactor pressure vessel (‘integral’ PWR design). All nuclear power reactor designs have, enhanced safety features relative to current LWR.
Building on the experience of several innovative reactors built in the 1960s and 1970s, new high-temperature gas-cooled reactors (HTR), being developed. These reactors will be capable of delivering high temperature (up to about 1000°C) helium either for industrial application via a heat exchanger, or to make steam conventionally via a steam generator, or directly to drive a Brayton cycle gas turbine for electricity with almost 50 percent thermal efficiency possible (efficiency increases around 1.5 percent with each 50°C increment). Improved metallurgy and technology developed in the last decade makes HTR more practical than in the past, though the direct cycle means that there must be high integrity of fuel and reactor components.
Fuel for these 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 percent 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.
HTR 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 HTR. With negative temperature coefficient of reactivity (the fission reaction slows as temperature increases) and passive decay heat removal, the reactors are inherently safe. HTR 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.
Here are some major designs, which are, covered under this category:
Fast neutron reactors have no moderator, a higher neutron flux and are normally, cooled by liquid metal such as sodium, lead, or lead-bismuth, with high conductivity and boiling point. They operate at or near atmospheric pressure and have passive safety features, (most have convection circulating the primary coolant). Automatic load following is, achieved due to the reactivity feedback—constrained coolant flow leads to higher core temperature, which slows the reaction. Primary coolant flow is by convection. They typically use boron carbide control rods.
Here are the major designs under this category:
- HYERION POWER MODULE
- STAR-LM, STAR H2, SSTAR
- KOREAN FAST REACTOR DESIGNS
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. U-235 fluoride was in molten sodium and zirconium fluorides at 860°C that flowed through a graphite moderator. There is now renewed interest in the concept in Japan, Russia, France and the USA, and one of the six Generation IV designs selected for further development is the molten salt reactor (MSR).
In the MSR, the fuel is a molten mixture of lithium and beryllium fluoride salts with dissolved enriched uranium, thorium or U-233 fluorides. The core consists of unclad graphite moderator arranged to allow the flow of salt at some 700°C and at low pressure. Heat is, transferred to a secondary salt circuit and thence to steam. It is not a fast neutron reactor, but with some moderation by the graphite is epithermal (intermediate neutron speed). The fission products dissolve in the salt and are removed continuously in an on-line reprocessing loop and replaced with Th-232 or U-238. Actinides remain in the reactor until they fission or are converted to higher actinides which do so. MSR have a negative temperature coefficient of reactivity, so will shut down as temperature increases beyond design limits.
Here is the description for each:
Aqueous homogeneous reactors (AHR) have the fuel mixed with the moderator as a liquid.
Typically, low-enriched uranium nitrate is in aqueous solution. About 30 AHR 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. However, corrosion problems and the propensity of water to decompose radiolytically (due to fission fragments) releasing gas bubbles have been design problems.
In 2008, the IAEA summarized:
- “The use of solution reactors for the production of medical isotopes is potentially advantageous because of their low cost, small critical mass, inherent passive safety, and simplified fuel handling, processing and purification characteristics. These advantages stem partly from the fluid nature of the fuel and partly from the homogeneous mixture of the fuel and moderator in that an aqueous homogeneous reactor combines the attributes of liquid fuel homogeneous reactors with those of water moderated heterogeneous reactors. If practical methods for handling a radioactive aqueous fuel system are implemented, the inherent simplicity of this type of reactor should result in considerable economic gains in the production of medical isotopes.”
Thermal power can be 50-300 MW at low temperature and pressure, and low enriched uranium fuel used. However, recovering desired isotopes on a continuous production basis remains to be, demonstrated. As well as those in solution, a number of volatile radioisotopes used in nuclear medicine can be, recovered from the off-gas arising from radiolytic “boiling”. For isotopes such a Sr-89, this is very much more efficient than alternative production methods.
At the end of 2007, Babcock & Wilcox (B&W) notified the US Nuclear Regulatory Commission that it intended to apply for a license to construct and operate a Medical Isotope Production System (MIPS) – an AHR system with low-enriched uranium in small 100-200 kW units for Mo-99 production. A single production facility could have four such reactors. B&W expects a five-year lead-time to first production. The fuel is, brought to criticality in a 200-litre vessel. As fission proceeds, the solution is circulated through an extraction facility to remove the Mo-99 and then back into the reactor vessel, which is at low temperature and pressure. In January 2009, B&W Technical Services Group signed an agreement with radiopharmaceutical and medical device supplier Covidien to develop technology for the MIPS.
The IRIS developers have outlined the economic case for modular construction of their design (about 330 MWe), and the argument applies similarly to other smaller units.
They point out that IRIS with its size and simple design is ideally suited for modular construction 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 expect that construction of the first IRIS unit will be, completed in three years, with subsequent reduction to only two years.
Site layouts have been developed with multiple single units or multiple twin units. In each case, units will be, constructed so that there is physical separation sufficient to allow construction of the next unit while the previous one is operating and generating revenue. In spite of this separation, the plant footprint can be very compact so that a site with three IRIS single modules providing 1000 MWe capacity is similar or smaller than one with a comparable total power single unit.
Eventually IRIS is, expected to have a capital cost and production cost comparable with larger plants. However, any small unit such as this will potentially have a funding profile and flexibility otherwise impossible with larger plants. As one module is finished and starts producing electricity, it will generate positive cash flow for the next module to be, built.
Westinghouse estimates that 1000 MWe delivered by three IRIS units built at three-year intervals financed at 10 percent 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.
Dr. Mir F. Ali