Advanced Reactors: Today's nuclear reactor technology is distinctly better than that represented by most of the world's operating plants, and the first advanced reactors are now in service in Japan.
Reactor suppliers in North America, Japan and Europe have nine new nuclear reactor designs either approved or at advanced stages of planning, and others at a research and development stage.
These incorporate safety improvements including features which will allow operators more time to remedy safety problems and which will provide greater assurance regarding containment of radioactivity in all circumstances. New plants will also be simpler to operate, inspect, maintain and repair, thus increasing their overall reliability and economy.
The new generation reactors:
have a standardised design for each type to expedite licensing, reduce capital cost and reduce construction time,
are simpler and more rugged in design, easier to operate and less vulnerable to operational upsets-
have higher availability and longer operating life,
will be economically competitive in a range of sizes,
further reduce the possibility of core melt accidents,
have higher burn-up to reduce fuel use and the amount of waste.
The greatest departure from most designs now operating is that many new generation nuclear plants will have more 'passive' safety features which rely on gravity, natural convection, etc, requiring no active controls or operational intervention to avoid accidents in the event of malfunction.
The new designs fall into two broad categories: evolutionary and developmental. The evolutionary designs are those which are basically new models of existing, proven designs. The developmental designs depart more significantly from today’s plants and require more testing and verification before large-scale deployment.
In USA, the Department of Energy (DOE) and the commercial nuclear industry have been developing three advanced reactor types. Two of the three are large (1300 MWe) "evolutionary" designs which build directly on the experience of operating light water reactors in the United States, Japan and Western Europe.
One is an advanced boiling water reactor (ABWR), two examples of which are in commercial operation in Japan. The other type is an advanced pressurized water reactor (System 80+), which is ready for commercialization.
Two System 80 reactors under construction in South Korea incorporate many of the design features of the System 80+. The US Nuclear Regulatory Commission (NRC) gave final design certification for these in 1997.
Another, more innovative US advanced reactor is smaller - about 600 megawatts - and has passive safety features. The AP-600 gained NRC final design certification in 1999.
These are the first such generic certifications to be issued and will be valid for 15 years. Following an exhaustive public process, it means that safety issues within the scope of the certified designs have been fully resolved and hence will not be open to legal challenge during licensing for particular plants.
Separate from the NRC process and beyond its immediate requirements, the US nuclear industry selected one standardized design in each category - the large ABWR and the medium-sized AP-600, for detailed first-of-a-kind engineering (FOAKE) work.
The US$ 200 million program was half funded by DOE. It means that prospective buyers can now gain firm information on construction costs and schedules.
Another US design, the Gas Turbine - Modular Helium Reactor, developed from an earlier design, has its fuel as particles coated by ceramic to enable high temperature operation.
It is cooled by helium which directly drives a gas turbine, and will be built as modules of 250-285 MWe each. The inert nature of the coolant and resistance of the fuel to melting make the concept attractive. It is being developed by an international partnership in Russia and may be used to burn ex-weapons plutonium.
South Africa's Pebble Bed Modular Reactor also has a direct-cycle gas turbine generator and is being developed by a consortium led by the utility Eskom, drawing on German expertise. Modules will be of 110 MWe and thermal efficiency about 42-50 percent .
Fuel consists of tennis ball sized pebbles of graphite moderator containing 8 percent enriched UO2 and coated with silicon carbide.
The 330,000 fuel pebbles recycle through the reactor continuously until they are expended, giving an average enrichment in the fuel load of 5-6 percent and burn-up of 80,000 MWdays per tonne U. Each unit will finally discharge about 19 tones of spent pebbles per year to ventilated on-site storage bins.
Construction cost (when in clusters of ten units) is expected to be US$ 1000/kW and generating cost 1.6 US cents/kWh. Eskom and the South African Industries Development Corporation hold 55 percent the project, with BNFL 20 percent and PECO Energy 10 percent. A prototype is due to be built in 2001 for commercial operation in 2005.
In Japan, the first two ABWRs have started operating, as noted above. In relation to PWRs, Mitsubishi has designed an advanced model which is simpler and combines active and passive cooling systems to greater effect. Design work on this 1400 MWe reactor continues and it will be the basis of the next generation of Japanese PWRs.
In Canada, the CANDU-9 (925-1300 MWe) is developed from an existing design but has flexible fuel requirements ranging from natural uranium through slightly-enriched uranium, recovered uranium from reprocessing spent PWR fuel, mixed oxide (U & Pu) fuel, direct use of spent PWR fuel, to thorium, and possibly burning military plutonium or actinides separated from reprocessed LWR waste. A two year regulatory review of the CANDU-9 design was completed in1997.
In Europe, under a joint venture with French and German utilities, Nuclear Power International is developing a large (up to 1750 MWe) European pressurised water reactor (EPR). This is an evolutionary design which has been confirmed as the new standard for France, meeting stringent new European safety criteria.
In Russia, two advanced reactor designs have been developed. The largest of these is the VVER-1000 model V-392, an evolutionary PWR with passive safety features. A smaller version is the VVER-640 (V-407 type), with Western control systems.
The first four of these are being built near St Petersburg and are expected to start up from 2002. Small floating nuclear power plants are also being developed.
Fast neutron reactors:
Fast neutron reactors are a different technology from those considered so far. They generate power from plutonium by much more fully utilizing the uranium-238 in the reactor fuel assembly, instead of needing just the fissile U-235 isotope used in most reactors. If they are designed to produce more plutonium than they consume, they are called Fast Breeder Reactors (FBR). If they are net consumers of plutonium they are sometimes called burners.
For many years the focus has been on the potential of this kind of reactor to produce more fuel than they consume, but today, with low uranium prices and the need to dispose of plutonium from military weapons stockpiles, the main interest is in their role as incinerators.
Several countries have research and development programs for Fast Breeder Reactors (FBR), which are, generically, Fast Neutron Reactors. Over 290 reactor-years of operating experience has been gained on this type of plant.
In the closed fuel cycle it can be seen that conventional reactors produce two "surplus" materials; plutonium (from neutron capture, separated in reprocessing) and depleted uranium (from enrichment).
The fast neutron reactor uses plutonium as its basic fuel while at the same time converting depleted (or natural) uranium, basically U-238, comprising a "fertile blanket" around the core, into fissile plutonium.
In other words it "burns" and can "breed" plutonium, as shown in. Depending on the design, it is possible to recover from reprocessing the spent fuel enough fissile plutonium for its own needs, with some left over for future breeder reactors or for use in conventional reactors.
Fast neutron reactors have a high thermal efficiency due to their high-temperature operation. Cooling is by liquid sodium. Although in many ways this is difficult to handle chemically, in some respects it is more benign overall than very high pressure water.
Experiments on a 19 year old UK breeder reactor before it was decommissioned in 1977 showed that the liquid sodium cooling system made it less sensitive to coolant failures than the more conventional very high pressure water and steam systems in light water reactors. More recent operating experience with large French and UK prototypes has confirmed this.
The fast breeder reactor has the potential for utilizing virtually all of the uranium produced from mining operations.
Overall about 60 times more energy can be extracted from the original uranium by the fast breeder cycle than can be produced by the current light water reactors operating in "open cycle".
This extremely high energy efficiency makes the breeder an attractive energy conversion system. However, high capital costs and an abundance of low cost uranium means that they are unlikely to be competitive for several decades, probably not much before 2050.
For this reason design work on the 1450 MWe European FBR was phased out in 1994, although research at the 1250 MWe French Superphenix FBR took place 1995-98.
Research continues on the Indian FBRs, to pave the way to greater use of thorium as a fuel, and Japan's Monju prototype commercial FBR was connected to the grid in August 1995 (but was then shut down due to a major sodium leak).
The Russian BN-600 fast breeder reactor has been supplying electricity to the grid since 1980 and has the best operating and production record of all Russia's nuclear power units.
The BN-350 FBR operated in Kazakhstan for about 25 years and about half of its output was used for water desalination. Russia plans to build special fast neutron reactors to utilise the plutonium from its military stockpiles. About 20 FBRs have already been operating, some since the 1950s.
Thorium cycle:
Near-breeder or thorium cycle reactors are similar to fast breeders in that a fertile material, naturally-occurring thorium-232, will absorb slow neutrons to become (indirectly) fissile uranium-233.
This will produce a chain reaction yielding heat while surplus neutrons convert more thorium to U-233. The technology is considered by some to be attractive because plutonium production is avoided, fairly abundant thorium is used as a fuel, and the efficiency of fuel use approaches that of the fast breeder reactor.
However, the amount of fissile uranium produced is not quite enough to sustain the reaction, hence the term "near-breeder" is generally used. Though a focus of interest for 30 years, there are no commercial outcomes in sight.
Source:U.I.C.EDU
© 2000 Mena Report (www.menareport.com)
