<sup>What is a Pebble-bed Modular Reactor (PBMR)? 

A PBMR is a high-temperature helium-cooled reactor using a direct cycle gas turbine. Essentially it is a nuclear plant which is inherently safe, presents lower-cost options and facilitates problem-free siting.</sup>

^{Overall, nuclear power currently accounts for 20 percent of electricity generation in the United States. It produces reliable electricity generation with no greenhouse gas emissions. In 2000, nuclear generation displaced roughly 180 million metric tons of carbon, 3.8 million tons of sulfur dioxide and 2.3 million tons of nitrogen oxide, assuming the displaced capacity was based on average fossil fuel generation. In evaluating the future impact on emissions, the replacement fuel for retiring nuclear plants is of key importance. Given technology costs and fuel prices expected over the next 20 years, they would likely be replaced by natural gas-fired, combined-cycle plants with relatively low emission rates, compared to coal plants. In the AEO2002, carbon emissions varied from 3 million tons lower to 6 million tons higher in the high and low nuclear life extension cases, relative to the Reference Case. Greater shifts in nuclear generation, either decreasing through increased retirements, or increasing through new construction, could have a bigger impact on future emissions.} 

^{How does a PBMR work?} 

^{The nuclear fuel is contained in balls with a diameter of 60 mm. About 400 000 of these fuel balls will lie within a graphite-lined silo that will be 10 m high and 3,5 m in diameter. Helium at a temperature of about 500 €C is introduced into the top of the reactor. After the gas passes between the fuel balls, it leaves at the bottom at a temperature of about 900 C. This gas passes through three turbines. The first two turbines drive compressors and the third the generator, from where the power emerges. At that stage the gas is about 600 C. It then goes into a recuperator where it loses excess energy and leaves at about 140 €C. A water-cooled precooler takes it down further to about 30 C. The gas is then repressurized in a turbo-compressor before moving back to the regenerator heat-exchanger, where it picks up the residual energy and goes back into the reactor. Spent fuel balls are passed pneumatically to large storage tanks at the base of the plant where there is enough storage capacity to store all spent fuel throughout the life of the plant. The tanks are also designed to hold the spent fuel for 40 to 50 years after shutdown. About 2,5-million fuel balls will be required over the 40-year life of a 100 MW reactor.}

 

What is Generation IV? At the beginning of 2002, 438 nuclear power reactors were in operation in 31 countries around the world, generating electricity for nearly 1 billion people. They account for approximately 17 percent of worldwide installed base load capacity for electricity generation and provide half or more of the electricity in a number of countries. These reactors are generating electricity in a reliable, environmentally safe and affordable manner without emitting noxious gases into the atmosphere. Concerns over energy resource availability, climate change, air quality, and energy security suggest an important role for nuclear power in future energy supplies. While the current Generation II and III nuclear power plant designs provide an economically, technically, and publicly acceptable electricity supply in many markets, further advances in nuclear energy system design can broaden the opportunities for the use of nuclear energy. To explore these opportunities, the U.S. Department of Energy's Office of Nuclear Energy, Science and Technology has engaged governments, industry, and the research community worldwide in a wide-ranging discussion on the development of next-generation nuclear energy systems known as "Generation IV". This has resulted in the formation of the Generation-IV International Forum (GIF), a group whose member countries are interested in jointly defining the future of nuclear energy research and development. In short, "Generation IV" refers to the development and demonstration of one or more Generation IV nuclear energy systems that offer advantages in the areas of economics, safety and reliability, sustainability, and could be deployed commercially by 2030. A Generation IV Technology Roadmap is being prepared by GIF member countries which will identify the six to eight most promising reactor system and fuel cycle concepts and the R&D necessary to advance these concepts for potential commercialization. The Roadmap was initiated in October 2000 and is scheduled for completion in September 2002. New Reactor Designs  Overview This issue paper summarizes nuclear reactor designs that are either available or anticipated to become available in the United States by 2030. Criteria for including reactors are: 1) participation in the U.S. Nuclear Regulatory Commission's certification or pre-certification programs or 2) inclusion under the international Generation IV International Forum (GIF) program for longer-term reactor development. The U.S. Department of Energy is among the sponsors of the GIF program. While no detailed technical description of particular reactor designs is included, such descriptions and schematics are available elsewhere and, when practical, are hyperlinked in the text. Reactor vendors who put forward new designs anticipate that their designs will meet commercial market needs including an affordable, competitive construction cost and the usually low operating costs of commercial nuclear reactors. Such views are not assessed, though a section does identify public discussion of efforts by the nuclear industry and the U.S. government to improve the industry's competitive position.1 Existing Reactor Designs and Design Categories There are now 104 fully licensed nuclear power reactors in the United States though only 103 are now operational.2 Because each of these reactors is fully licensed and meets national safety standards, a potential builder might replicate any of these designs for future construction. This is less likely, however, because existing, operable reactors in the United States were licensed during or before the 1970s. Technology has progressed and any future construction should incorporate more advanced designs that better meet today's commercial and safety criteria. There are possible exceptions to the preceding statement. Three or four reactors in the United States were partially built and still possess valid construction licenses. These reactors are WNP-1 (Washington State), Watt's Bar 2 (Tennessee), and Bellefonte 1 and 2 (Atlanta). Moreover, these construction licenses have recently been extended to approximately the end of the present decade. Construction on each unit was halted over a decade and a half ago. Builders of these units, subject to the rules of their licenses, have the right to resume construction on their reactors, units that were designed during the 1970s or earlier. Whether the construction will resume and whether former designs will be continued remains to be determined. The owners of WNP-1 have announced plans to forgo their construction license to allow for disassembly. All existing commercial nuclear reactors operating in the United States fall into two broad categories, pressurized water reactor (PWR) and boiling water reactor (BWR). Because both types of reactors are cooled and moderated3 with ordinary "light" water, the two designs are often grouped collectively as light water reactors (LWR). LWRs generate power through steam turbines similar to those used for most power generated by burning coal or fuel oil. Light water reactors have so far proven to be the most commercially popular reactor design worldwide though there are notable exceptions.4  There are several available websites that discuss existing reactors in the United States. These include http://www.eia.doe.gov/cneaf/nuclear/page/nuc_reactors/reactsum.html. Information on international operating reactors is available at http://www.iaea.org/programmes/a2. PWRs use nuclear-fission to heat water under pressure within the reactor. This water is then sent to a heat exchanger (called a "steam generator" in PWRs) where steam is produced to drive an electric generator. The water used as a coolant in the reactor and the water used to provide steam to the electric turbines exists in separate closed loops that involve no discharges to the environment. Of the 104 fully licensed reactors in the United States, 69 are PWRs. Westinghouse, Babcock and Wilcox, and Combustion Engineering designed the units operating in the U.S. After these reactors were built, Westinghouse and Combustion Engineering nuclear assets were combined with British Nuclear Fuels Limited to form Westinghouse BNFL. The French-German firm Framatome ANP has acquired many of Babcock and Wilcox's nuclear technology rights, though portions of the original Babcock and Wilcox firm still exist and also possess some technology rights as well. Other major makers of PWR reactors, including Framatome ANP and the Russian Atomstroyexport, have not yet sold their reactors in the U.S. A schematic diagram of a PWR can be found at http://www.eia.doe.gov/cneaf/nuclear/page/nuc_reactors/pwr.html. The remaining 35 operable commercial nuclear reactors in the United States are BWRs. BWRs allow fission-based heat from the reactor core to boil the coolant water directly into the steam (i.e. no heat transfer) that is used to generate electricity. General Electric built all boiling water reactors now operational in the United States. Framatome ANP and Westinghouse BNFL have each designed BWRs though these have not yet been sold in the United States. A schematic diagram of a BWR can be found at http://www.eia.doe.gov/cneaf/nuclear/page/nuc_reactors/bwr.html. Although no LWR projects have been initiated in the United States since the 1970s, the overall performance record of the existing fleet has been reasonably successful. Some 111 LWRs have entered service in the U.S. since 1969.5 Only seven of those since 1969 have been permanently shut down. The average annual capacity factor for nuclear reactors in the United States has been around 90 percent during the early 2000's. Average operating costs, as reported by the Federal Energy Regulatory Commission, are slightly lower for LWRs than for operating coal-fired plants and considerably below operating costs for gas-fired plants.6 There have been attempts to operate additional classes of reactors in the United States, though most examples were prototypes and were not commercial successes. Perhaps the most famous example was the Fort Saint Vrain reactor that operated between 1974 and 1989. It was a high temperature gas-cooled reactor or HTGR. Other HTGRs operated elsewhere notably in Germany. HTGRs, of which there are many sub-categories, continue to stimulate commercial interest. Small prototypes now operate in China and Japan and additional HTGR designs are promoted by firms in South Africa, the United States, the Netherlands, and France. HTGRs use a gas- recently helium has been preferred- to generate electricity. In some cases the turbine is run directly by the gas, in other cases steam or alternative hot gases are produced in a heat exchanger to generate the power. HTGRs are distinguished from other gas-cooled reactors by the higher temperatures attained within the reactor. Such higher temperatures might permit the reactor to be used as an industrial heat source in addition to generating electricity. This improves their suitability for hydrogen production. Advocates of HTGR designs hold that HTGR's have high safety, low costs, and an ability to supply power to smaller markets. Commercial reactor designs that operate outside of the United States include fast breeder reactors (FBRs), pressurized heavy water reactors (PHWRs), and gas-cooled reactors (GCRs). FBRs have received only very limited market support, though "commercial" units operate in Russia and France, and prototypes exist elsewhere, notably Japan and India. "Breeder" or "fast" reactors have advantages because U-235 is the only naturally occurring uranium isotope that is directly suitable for commercial energy production. U-235 is only 0.7 percent of natural uranium.7 Most of natural uranium is the U-238 isotope that is not directly usable as a reactor fuel. During the course of any reactor operation a portion of the U-238 in the fuel is converted to plutonium, primarily the useful Pu-239 isotope, which provides much of the energy used in nuclear power production. The bulk of the U-238 content in an LWR is not converted to plutonium and the unconverted U-238 does not contribute significantly to power generation. A breeder reactor converts more U-238 to usable fuels than the reactor consumes. Any unused fuel would have to be "reprocessed" before some of the plutonium and the remaining U-238 would again be usable. FBRs have, so far, proven to be more expensive to build and operate than LWRs. It is not yet clear whether this is due to the fact that most FBRs have been prototypes or if this reflects underlying costs. The plutonium content of the reprocessed fuel also raises concerns over weapons proliferation. Many early FBR designs were prone to system failures, though some, notably the BN-600 unit in Russia, have operated over extended periods. Proponents of advanced reactor designs believe that some commercial FBR designs could be deployed prior to many other advanced, but untested commercial designs.8 PHWRs have been promoted primarily in Canada and India, with additional commercial operating units in several other nations including South Korea, China, Romania, and Argentina. Canadian designed PHWRs are often called "CANDU" reactors.9 Siemens, ABB, and Indian firms have also built commercial PHWR reactors. Commercial heavy water reactors now in operation use heavy water as moderators and coolants. No successful effort has been made to license PHWRs in the United States. PHWRs have proven to be popular in several countries because they usually use less expensive natural (not enriched) uranium fuels and can be built and operated at competitive costs. PHWRs have often been preferred by nations wishing to develop an indigenous fuel cycle without expensive enrichment facilities. Proliferation issues related to the continuous process of refueling PHWRs have raised some concerns as has spent fuel composition.10 The term gas-cooled reactor (GCR) can be used ambiguously . HTGRs, for example, are clearly a subset of GCRs that operate at higher temperatures. As used here, GCRs are "Magnox" reactors designed and built in the United Kingdom since the 1950s and the derivative, advanced gas-cooled reactor (AGR), also operated in the United Kingdom. Similar reactors have been built in France, Sweden, and Japan. No GCR design has operated commercially in the United States. Existing GCR designs have not been commercially successful outside of the United Kingdom. Commercial GCRs11 in the United Kingdom have operated longer than commercial reactors anywhere else in the world. Like the PHWRs, the original GCR designs use natural uranium fuels, though newer designs use slightly enriched fuels and are not confined to uranium. 12 Other potential designs for commercial reactors abound. They have not been widely considered in recent history in the United States. There is some experience with additional concepts elsewhere. New Designs 1. Certified Designs Following statutory requirements, the Nuclear Regulatory Commission (NRC) has set up a process by which reactor designs might be certified prior to any actual construction plans. The certification process seeks to reduce site development time by resolving common design issues prior to construction. Design certification is an optional process and may occur simultaneously with site licensing. Certification Process for New Reactors in the United States, Design, Type and Present State Reactor Design Lead Vender(s) Design Category Status at NRC System 80+ Westinghouse BNFL PWR Certified ABWR GE, Toshiba, Hitachi BWR Certified AP600 Westinghouse BNFL PWR Certified AP1000 Westinghouse BNFL PWR Certification ESBWR GE BWR Pre-certification SWR-1000 Framatome ANP BWR Pre-certification ACR-700 AECL PHWR/PWR hybrid Pre-certification PBMR Eskom HTGR Pre-certification GT-MHR General Atomic HTGR Pre-certification IRIS Westinghouse BNFL PWR Pre-certification EPR Framatome ANP PWR No application decision ACR-1000 AECL PHWR No application decision Note: Reactor names are defined in the text. Any new reactor built in the United States over the next decade or so would probably use designs either recently certified by the NRC or that will be certified by the NRC in the near future. The re-creation of older designs is popular overseas and cannot be ruled out in the United States. Presently there are three certified new reactor designs in the United States: the System 80+, the Advanced Boiling Water Reactors (ABWR), and the AP600. These designs are sometimes called Advanced Light Water Reactors (ALWR) because they incorporate more advanced safety concepts than the reactors previously offered by vendors. They are also sometimes called Generation III reactors to distinguish them from earlier designs now operating in the US.  System 80+ (Westinghouse BNFL): The System 80+ reactor is a PWR that was designed by Combustion Engineering and by CE's successor owners ABB and Westinghouse BNFL. The NRC has certified the System 80+ for the U.S. market, but Westinghouse BNFL no longer actively promotes the design for domestic sale. The System 80+ provides the basis for the APR1400 that has been developed in Korea for future deployment. I ABWR (General Electric, Toshiba, Hitachi): Of the three NRC- certified ALWR designs only the ABWR has been deployed. Three ABWRs operate in Japan, and four are under construction, two each in Taiwan and Japan. While the ABWR design is usually associated in the United States with General Electric, the units now being built in Japan are products of Toshiba and Hitachi. General Electric, Toshiba, and Hitachi have shown an interest in building ABWRs in the U.S. There are many variations in ABWR design. The most frequently mentioned capacities in the 1250-1500 MWe range. Smaller and larger designs exist depending on vendor. Vendors now claim costs for the ABWR that have attracted some customer interest.  AP600 (Westinghouse BNFL): The AP600 is a PWR designed by Westinghouse BNFL and certified by the NRC. The AP600, while based on previous designs, has innovative passive safety features that permit a greatly simplified reactor design. Simplification has reduced plant components and construction costs. The AP600 has been bid overseas but has never been built. Westinghouse has recently de-emphasized the AP600 in favor of the larger, though potentially less expensive AP1000 design.  The initial ALWR reactors as a group have been praised for their improvements in reactor safety and simplicity, but construction costs on a per kilowatt of capacity basis might still be a barrier to commercial success in the U.S. The ABWR design however has many variations and continues to be promoted in the U.S. by several vendors. It is being considered for construction at Bellefonte by the Tennessee Valley Authority (TVA). 2. Undergoing Certification Only one reactor design is presently undergoing certification with the NRC, although this situation could change shortly as additional designs move from "pre-certification" to actual "certification". The process of certification can take several years and depends heavily on what design is proposed and supported by potential vendors and buyers. AP1000 (Westinghouse BNFL): Quite often when a reactor is named, its name includes digits such as the "1000" in the AP1000. This usually indicates the initial electricity generating capacity of the design, in this case 1000 MWe. Seldom do do the digits mean the present capacity of the design. The most recent AP1000 now has 1117 MWe capacity. The AP1000 is an enlargement of the initial AP600, designed to increase the reactor's target output by about 90 percent without significantly increasing the total cost of building the reactor. Operating costs are anticipated to be less than the AP600. While Westinghouse BNFL owns rights to several other designs, the AP1000 is the principal product that the company now promotes in the United States. The AP1000 is a PWR with innovative, passive safety features and a much simplified design that is intended to cut the material and construction costs of the plant.One consortium of 9 utilities called NuStart Energy is promoting the AP1000 design.  3. Undergoing Pre-Certification While pre-certification is a technical concept within the NRC regulatory environment, the process can mean many things to potential reactor manufacturers. Concepts such as the ESBWR, the SWR-1000, and the ACR-700 appear to be much further along toward potential deployment than, say, the IRIS and GT-MHR designs.13 Pre-certification, however, represents a vendor's intention to proceed toward commercialization in the U.S. and perhaps globally. Pre-certification is a less expensive stage of the overall certification process. Actual certification procedures are much more complex and many NRC costs are born by the applicant.  ESBWR (General Electric): The ESBWR14 is a new simplified BWR design being promoted by General Electric. It constitutes an evolution and merging of several earlier design ideas including the ABWR and other designs that are no longer being actively pursued by GE. The intent of the new design, which includes new passive safety features, is to cut construction and operating costs significantly from the ABWR design. GE is investing heavily in the ESBWR though the design might not be available for deployment for several years. The nine utility NuStart Energy group promotes the ESBWR as well as the AP1000 design.  Siedewasser Reaktor (SWR-1000) (Framatome ANP): The SWR-1000 is a Framatome ANP design for an advanced BWR. Framatome ANP was created through the merger of the French nuclear vendor Framatome and the nuclear power assets of the German firm Siemens. The SWR-1000 was originally designed by Siemens. Framatome ANP has also recently begun SWR-1000 pre-certification with the NRC. Literature on the design emphasizes the reactor's passive safety features. Passive safety should also mean lower construction costs though this is not emphasized by Framatome. Information on the SWR1000 can be found on http://www.de.framatome-anp.com/anp/e/foa/anp/products/s112.htm. The SWR-1000 presently has no US utility sponsor. Information related to certification of the SWR-1000 can be found at http://www.nrc.gov/reactors/new-licensing/license-reviews/swr-1000.html.   ACR-700 (Atomic Energy of Canada Limited): AECL's "Advanced CANDU Reactor" ACR-70015 has been developed over a lengthy period of time and is considered an evolution from AECL's internationally successful CANDU line of PHWRs. CANDU reactors have been more of a commercial success than any other line of power reactors other than the LWRs. One of the innovations in the ACR-700, compared to earlier CANDU designs, is that heavy water is used only as a moderator in the reactor. Light water is used as the coolant. Earlier CANDU designs used heavy water both as a moderator and as a coolant. This change makes it debatable whether the ACR-700 is a PHWR, a PWR, or a hybrid between the two designs. This AECL has aggressively marketed the ACR-700 offering low prices, short construction periods, and favorable financial terms. As is the case for most non-LWR reactors, most U.S. utilities, nuclear engineers, and regulators have only limited working familiarity with the design. Interest has been shown by Dominion Resrouces regarding possible ACR-700 construction at North Anna (Virginia) and in Canada by Canadian firms.  Pebble-bed Modular Reactor  (PBMR) The PBMR, which uses helium as a coolant, is part of the HTGR family of reactors and thus a product of a lengthy history of research, notably in Germany. More recently the design promoted and revised by the South African utility Eskom. Eskom continues to partner in the design with BNFL among its' investors. Recently Eskom is expected to receive approval to build a prototype PBMR in South Africa. Certification procedures in the U.S. have slowed but never has been abandoned. At around 165 MWe the PBMR is one of the smallest reactors now being proposed for the commercial market. This is considered a marketing advantage because new small units require less capital investments than larger new units. Small size has been viewed as a regulatory disadvantage because most licensing regulations (at least formerly) required separate licenses for each unit at a site. Fuels used in the PBMR would be more highly enriched uranium than is presently used in LWR designs. The design is considered a possible contender for the US Department of Energy's Next Generation Nuclear Plan (NGNP) program in Idaho.  Information related to certification of the PBMR can be found at http://www.nrc.gov/reactors/new-licensing/license-reviews/pbmr.html. Gas-turbine Modular Helium Reactor (GT-MHR) (General Atomic): The GT-MHR is an HTGR design that has been developed primarily by the U.S. firm, General Atomics. The most advanced plans for GT-MHR development relate to building reactors in Russia to assist in the "burn up" of surplus plutonium supplies. Parallel plans for commercial power reactors would use uranium based-fuels enriched to as high as 19.9 percent U-235 content. This would keep the fuel below the 20 percent enrichment level that defines highly enriched uranium. In initial designs, the conversion of the energy in the heated helium coolant to electricity would be directly in a gas turbine. There has been concern regarding untested aspects of this technology. This has led some potential sponsors to propose less innovative heat transfer mechanisms to generate electricity. The U.S. utility, Entergy, has participated in GT-MHR development and has used the name "Freedom Reactor" for the design. Because coolant temperatures arising from HTGR reactors are much higher than from LWRs the design is viewed as a potential commercial heat source. There has been particular attention to the design's potential in a non-polluting method to produce hydrogen. The GT-MHR is considered a potential contender for the US Department of Energy's Next Generation Nuclear Plan (NGNP) program in Idaho. Information on the GT-MHR can be found on http://www.ga.com/gtmhr/. Information related to certification of the GT-MHR can be found at http://www.nrc.gov/reactors/new-licensing/license-reviews/gt-mhr.html. International Reactor Innovative & Secure (IRIS) (Westinghouse BNFL): Westinghouse BNFL has promoted the IRIS reactor design as a significant simplification and innovation in PWR design. The reactor would be smaller than most operating PWRs and would be much simplified. The IRIS reactor includes features designs that are intended to avoid loss of coolant accidents. Research has continued for some time on the design and pre-certification is in process. The IRIS is viewed as not ready for development during the present decade, but may show potential during the next decade. IRIS has a targeted 2015 completion date for the design. The design presently has not utility sponsor but this is to anticipate early commercial interest.Information on the IRIS can be found on http://www.nei.org/index.asp?catnum=3&catid=712 and through http://www.nrc.gov/reactors/new-licensing/new-licensing-files/ml030780800.pdf. Information related to certification of the IRIS can be found at http://www.nrc.gov/reactors/new-licensing/license-reviews/iris.html. 4. Anticipated for Possible Pre-Certification Two designs, the European Pressurized Water Reactor (EPR) and the ACR-1000, have not been submitted for pre-certification in the United States. Because of the attention that the designs are now receiving, they are described below. EPR (Framatome ANP): Framatome ANP has not decided if it will market its EPR in the United States. The EPR is a rather conventional PWR unit though components have been simplified and considerable emphasis is placed on reactor safety. The design is being built in Finland. Additionally, the French government has proposed building an EPR in France. Nuclear power already constitutes over 75 percent of France's power supply; therefore building a new reactor in France might lead to decommissioning an existing reactor to make room in the market for the base load power provided by an EPR. France has proposed the EPR construction to China. The proposed size of the reactor would be around 1600 MWe though earlier designs were as large as 1750 MWe. Framatome has indicated that it might seek US certification of the EPR after European development proceeds.  ACR-1000 (Atomic Energy of Canada Limited): While AECL promotes its ACR-700 design, an ACR-1000 is being designed as well. If the scale economies attributed by Westinghouse BNFL to its AP series and to GE's ABWR are valid, one might anticipate parallel, cost-lowering results for the ACR series. Advertised costs for the ACR-700 are already as low as any design proposed for the near term. Promised construction times of three years would set modern records for larger reactors. Information on the ACR-1000 can be found on http://www.aecl.ca/index.asp. 5. Generation IV Concepts The U.S. Department of Energy participates in the Generation IV International Forum (GIF), an association of twelve nations that seek to develop a new generation of commercial nuclear reactor designs before 2030. Criteria for inclusion of a reactor design for consideration by the GIF group include:        1. Sustainable energy (extended fuel availability, positive environmental impact)        2. Competitive energy (low costs, short construction times)        3. Safe and reliable systems (inherent safety features, public confidence in nuclear            energy safety)        4. Proliferation resistance (does not add unduly to unsecured nuclear material) and            physical protection; (secure from terrorist attacks) During 2002, GIF members agreed to concentrate their efforts and funds on six concept designs that they seek to become commercially viable between 2015 and 2025. There is thus some leeway between the 2030 target for the GIF program and the targets for individual concepts. Individual GIF participant nations are free to pursue any technology they chose. The United States intends to pursue each design.. The GIF group, along with the U.S. Department of Energy's Nuclear Energy Research Advisory Committee (NERAC), has published "A Technological Roadmap for Generation IV Nuclear Energy Systems" (December 2002) which summarizes plans and designs for Generation IV projects. This is accessible through http://www.nuclear.gov/ and describes each design in some detail including reactor schematics. Each design is evolutionary; thus while the following descriptions involve comparisons, these analogies should be interpreted with caution. Gen IV programs are summarized on http://www.inel.gov/initiatives/generation.shtml.   Gas-cooled Fast Reactor (GFR): The GFR uses helium coolant directly to a gas turbine generator to produce electricity. This parallels PBMR and original GT-MHR designs. The primary difference from these designs is that the GFR would be a "fast", or breeder reactor. One favored aspect of the design is that it would minimize the production of many undesirable spent fuel waste streams. The reference design size is targeted to be 288 MWe with a deployment target date of 2025. In addition to producing electricity the design might be used as a process heat source in the production of hydrogen. Lead-cooled Fast Reactor (LFR): So far, most breeder reactors have used molten metal technologies for their coolants. Many FBRs have used molten sodium, a metal with which there is considerable experience but which has sometimes been difficult to handle.The LFR uses molten lead or a lead-bismuth alloy as its coolant. One design favored under the Generation IV would result in long periods between refuelings, 15-20 years. Similar designs have been investigated in Russia. Target ranges for this reactor would be 50-150 MWe. That would be rather small by historic nuclear standards, but might meet localized market needs. Designs as large as 1200 MWe have been suggested. Initial targeted deployment would be in 2025. Proposed designs would favor electricity production though proponents consider the production of process heat at LFRs as possible.  Molten Salt Reactor (MSR): The MSR involves a circulating liquid of sodium, zirconium, and uranium fluorides as a reactor fuel. The MSR has been presented as providing a comparatively thorough fuel burn, safe operation, and proliferation resistance. The initial reference design would be 1000 MWe with a deployment target date of 2025. The design could use a wide variety of fuel cycles. Temperatures for electricity production would not be as hot as for some other advanced reactors but some process heat potential exists. Versions of the MSR have been around for some time but never were implemented for commercial uses. During 2003, the MSR was down rated within the Gen IV program because it was seen as too distant into the future for inclusion within the Gen IV schedule. Sodium-cooled Fast Reactor (SFR): Sodium-cooled fast reactors have been the most popular design for breeder reactors. Designs have been proposed under the "Technological Roadmap" ranging from 150 to 1700 MWe. Molten metal technology is no longer "new" but several early SFR prototypes had difficulty obtaining sustained operation. The BN-600 in Russia has been regarded as highly reliable. Design supporters believe that the SFR promises superior fuel management characteristics. The original target deployment date of 2015 reflects the considerable research that the design has already received. This date seems to be lagging as the VHTR gains favor. Earlier prototypes have already been built in France, Japan, Germany, the United Kingdom, Russia, and the United States since as early as 1951. Initial deployment would probably focus on electricity due to comparatively low "outlet temperatures" for the design. Sodium cooled reactors are discussed at  Supercritical-water-cooled Reactor (SCWR): The SCWR design is to be the next step in LWR development and has been proposed with alternatives that evolve from the BWR and PWR. SCWRs would operate at higher temperatures and thermal efficiencies than present LWRs. The reference plant would be 1700 MWe, toward the upper end of present LWR designs. The deployment target date is 2025. Some GIF participants favor the design. Most research on the design has been in Japan. Designers intend the SCWR to be much less expensive to build than today's LWR units though some of the economies appear to be shared by units now undergoing certification. Operating cost savings are also anticipated. Very-high-temperature Reactor (VHTR): The VHTR is an evolution from the HTGR family of reactors but would operate at even higher temperatures than designs now undergoing pre-certification. In contrast with the GFR, the VHTR would not be a breeder reactor, thus it would produce less potentially usable fuel than it consumes. In addition to generating electricity, the design would provide process heat that could be used in industrial activities including hydrogen production and desalinization. Electricity generation targets have not yet been set. Deployment is targeted for 2020, earlier than most other Generation IV designs. The VHTR is now the favored design in the US, where it is the basis for the Next Generation Nuclear Plant (NGNP) program in Idaho. France also favors the design.  Each GIF project involves new or untested concepts in reactor design. It would be surprising if every design concept met the program's initial targets. The research involved in the program has the potential to contribute to the understanding of alternative types of commercial nuclear power and process heat even if individual projects do not meet expectations. 6. Outlook Efficiency Issues A primary source of doubt regarding the potential of nuclear power, at least in the U.S., has been whether the recent technology has been too expensive to compete in the commercial marketplace. There have been relatively few orders for new nuclear power plants during the last two decades, not just in the United States and Canada, but also in Western Europe. Interest in new nuclear power units has recently focused on Asia and to a lesser extent in Eastern Europe. New orders in Finland and potentially France follow a long period of market inactivity. Reactor vendors have not ignored the message that their product has recently involved high construction costs and long construction periods. Vendors are attempting to position their product with promises of lower prices, shorter construction times, and specified financial arrangements. Most competitors are now offering fixed and historically low prices for their designs, though such prices are often confined to those parts of construction that the vendors actually control, the "nuclear island" that they designed. Concerns regarding construction costs for new nuclear power plants contrast sharply with the comparatively low cost of operating commercial reactor designs. Overall operating costs for nuclear power plants, as reported by the Federal Energy Regulatory Commission (FERC), have been roughly the same as and most recently slightly less than operating costs for coal-fired plants for about two decades. Such operating costs are considerably below the costs of operating most gas-fired generation units. Moreover, the fuel cost component of operating a nuclear power plant is particularly low. This operating cost advantage has given existing nuclear power units a favored position in the provision of base load electric power. Nuclear plant designers hope to take advantage of such low operating costs in positioning their new designs. Discussions of estimates of the capital and operating cost of new power generation units can be found on in the "Issues in Focus" section of the Annual Energy Outlook and in the Electricity Section of the Assumptions for the Annual Energy Outlook. The following publications summarize efforts and procedures to make new nuclear power plants commercially attractive. "Strategies for competitive nuclear power plants (TECDOC-1123)" International Atomic Energy Agency  (November 1999), website: http://www.iaea.org/worldatom "A Roadmap to Deploy New Nuclear Power Plants in the United States by 2010," United States Department of Energy Office of Nuclear Energy, Science and Technology and its Nuclear Energy Research Advisory Committee Subcommittee on Generation IV Technology Planning (October 31,  2001), website: http://www.nuclear.gov/  Scully Capital, "Final Draft, Business Case for Nuclear Power Plants, Bringing Public and Private Resources Together for Nuclear Energy" (July 2002) (available through United States Department of   Energy Office of Nuclear Energy, Science and Technology, website: http://www.nuclear.gov/ "A Technology Roadmap for Generation IV Nuclear Energy Systems (GIF-002-00)" U.S. DOE Nuclear  Energy Research Advisory Committee and Generation IV International Forum (December  2002),  website: http://www.nuclear.gov/ Summary and Potential There are early signs that the nature of the nuclear reactor market might be changing. Finland is building based on Framatome ANP's EPR design for a "fifth nuclear reactor". France also plans to build an EPR. More recently Bulgaria has discussed building a new nuclear reactor at Belene. Belene was originally to be a Russian-designed VVER-1000 unit using equipment already owned by and located in Bulgaria. Competitive submissions from several vendors for alternative, newer designs have been considered. It is unclear if the original design plan for Belene will move forward or if new designs will be slated. The United States is funding a program called Nuclear Power 2010 that seeks to build at least two nuclear power reactors by the mid 2010's. Supporting this has been proposed Federal energy legislation. Meeting the target would be a challenging task and the proposed legislation is still being debated .  1A large number of reactor designs have been excluded from the discussion. These include reactors promoted overseas by nations such as Russia, India, Argentina, Korea, Canada, and China, as well as numerous smaller or even portable reactors that are being examined worldwide, including in the United States. Also excluded is the International Atomic Energy Agency's International Project on Innovative Nuclear Reactors and Fuel Programs (INPRO) that covers territory similar to the GIF program in addition to other promising designs. GIF designs have been more heavily promoted within the United States.  2The one that is not operational, Brown's Ferry 1, has been shut down since 1985, but has not given up its operating license. The plant's owner-operator, the Tennessee Valley Authority, intends to restart the reactor in 2007. 3The terms "cooled" and "moderated" are important because they define reactor categories. Cooling in a reactor refers to the process and medium by which heat is transferred from the reactor core to the steam supply cycle of the nuclear power plant. "Moderating" is a concept unique to nuclear power. A moderator controls the rate of the nuclear power reaction and thus the amount of heat that is generated. In a light water reactor ordinary water serves both functions. Light water contains the same isotopes of hydrogen and oxygen as naturally occurring water. Heavy water contains a different, heavier isotope of hydrogen known as deuterium. Beyond the point that these conditions define reactor types, this will not matter in the discussion of existing reactors. It does matter for the group that will be discussed under "Generation 4" reactors. 4Exceptions include Canada, the United Kingdom, India, and part of Russia's industry. 5 Prior to 1969, some smaller commercial reactors were placed in service. All have been retired. 6 This is based on Utility Data Institute/Resource Data Internationl Compilations of FERC Form 1 data. 7The discussion here does not directly address "enrichment" the process by which the U-235 content of nuclear fuel is increased. 8This latter statement is based on "A Technology Roadmap for Generation IV Nuclear Energy Systems". 9Candu is a contraction of the term "Canadium deuterium". Canada has an interesting and unique nuclear power history which is covered by the book, Atomic Energy of Canada Limited, Canada Enters the Nuclear Age. 10Inspectors of nuclear power plants have a preference for plants such as the LWRs that are refueled in batches rather than the continuous fueling of PHWRs. Batch refueling is more easily monitored and occurs at intervals of one to two years. 11 Not the AGRs. 12 Most designs of PHWRs also use natural uranium fuels. However, variations in fuel type are possible at any PHWR with plutonium and thorium fuel content subject to particular interest and experimentation. 13 This sentence is a good example of the acronyms that overwhelm the nuclear steam supply system (NSSS) industry. Several of these acronyms no longer have any meaning in "words" while others have only limited actual meaning. They are defined below when possible. 14 The term ESBWR is now called the "EconomicSimplified Boiling Water Reactor". Definitions of the initials have changed overtime. 15 ACR is usually read to mean "Advanced CANDU Reactor". The Principles of Nuclear Power In naturally occurring uranium, 0.7% of uranium is of a particular type (isotope) of uranium (U235) which spontaneously splits (fissile material) to emit a tiny particle (a neutron). If this neutron hits another U235 atom, it too will split (a fission) to produce two more neutrons (chain reaction). If the concentration of U235 is sufficient (a critical mass), the process will be self-sustaining (the plant is `critical'), producing large quantities of heat in the `core' of the reactor.   Two important ingredients are needed to control the process and to utilise the heat, the moderator and the coolant. A moderator is a substance which neutrons collide with but `bounce off' without absorbing too much energy and without itself being split. It controls the amount of neutrons escaping from the core before they have hit another U235 atom. A good moderator is one which absorbs the least energy and does not absorb the neutrons before they split another uranium atom. Graphite is an excellent moderator; ordinary water is a poorer moderator but is much cheaper. If water is used, the U235 content must be increased (enrichment) to about 3 per cent to allow a chain reaction to take place. A rare isotope of hydrogen (deuterium) can be used to make so-called heavy water (deuterium is twice the weight of normal hydrogen) and this is also an excellent moderator.   In so-called fast (breeder) reactors (as opposed to the thermal reactors described above), no moderator is used and some of the neutrons escape the core and strike a `jacket' of uranium where they convert the unused part of the uranium, U238, to fissile material, plutonium, which can be used as a reactor fuel. The jacket is processed to isolate the plutonium for use in more fast reactors. The attraction of this design is obvious, it can use almost 100 per cent of naturally occurring uranium instead of the 0.7 per cent thermal reactors achieve. The disadvantage is equally obvious: it requires the separation, transport and widespread use of the material used to make nearly all nuclear weapons and is regarded as a serious proliferation risk. The technical attractions of the design have lead to huge amounts of public money being spent on this technology. However, in practice, all prototype plants have proved most unreliable and the technology is now all but abandoned.   In order to produce electricity, the heat in the core has to be transferred to a fluid (a liquid or a gas), the coolant. The heat will expand the fluid (boil it if it is water) and the force of the expanding gas can be used to drive a turbine generator to produce electricity. This principle of transferring heat from a `boiler' to a turbine generator is the same for all types of thermal power station whether it uses nuclear or fossil fuel. The coolant can go directly from the core to the turbine generator or there can be an intermediate stage where the coolant goes through a heat exchanger to produce steam in a second circuit. Liquids are much denser than gases and so a given volume of liquid can cool much more efficiently than the same volume of gas, so if the coolant circuit with a liquid cooled reactor breaks, the plant will only be cooled by gases, that is, steam and air, and the plant could over-heat catastrophically.   Ordinary water is a common, cheap coolant for power plants of all types, including nuclear power. Its primary safety disadvantage in a nuclear power plant is that if it escapes, the reactor will not be properly cooled (loss of coolant accident, or LOCA). Water can also be corrosive and will require expensive materials to prevent damage to the coolant pipes. However, water coolant requires much less volume of materials because of its greater efficiency in cooling than gas. So pressurized water reactors (PWRs) of the type built at Koeberg in South Africa, which use water as the coolant, are much more compact than, for example, the British designs of gas-cooled reactor. Of the gas coolants possible, carbon dioxide was used in the British power plant designs, but while this is cheap, it is somewhat corrosive. Helium is entirely inert, but is expensive so leakage has to be avoided.   Of the many possible technologies, two are of particular relevance to South Africa, the two existing civil nuclear power reactors at Koeberg and the PBMR. The Koeberg plants are each 900 MW (1 megawatt (MW) is 1 million kilowatts (kW)). They are known as pressurised water reactors (PWRs) because the coolant is maintained as liquid despite being at about 300°C by keeping it at very high pressures. This coolant is passed through a heat exchanger in which the energy is transferred to a second circuit in which water is boiled and drives the steam turbine generator.    Ordinary water is used as the moderator and as a result, uranium enriched to about 3 per cent is required.   The PWR is the most widely used design of nuclear reactor in the world and just under half the 430 nuclear power plants in the world are of this design. The main supplier is Westinghouse and its design has been adopted by Framatome (the Koeberg supplier), Siemens and Mitsubishi. The PWR is a direct descendant of submarine propulsion units and, as a result, its operating schedule is planned around annual stoppages when the plant is refuelled and maintenance is carried out. Typically, a quarter of the fuel rods are replaced each year, because the concentration of U235 is no longer great enough to maintain full power operation.   The PBMR uses helium as the coolant and graphite as the moderator and is one of a number of designs that come under the general classification of High Temperature (Gas-Cooled) Reactors, HTGRs or HTRs. The use of helium and graphite gives it several intrinsic safety and technical advantages over, say, the PWR. As noted above, the use of a gaseous coolant reduces the risk from loss of coolant accidents. Being inert, helium can be used at very high temperatures without concerns about corrosion.   The use of a good moderator like graphite increases the efficiency with which the uranium is used. With HTRs, fuel is made in ceramic pellets (or pebbles) which can also withstand very high temperatures, compared to a PWR where the fuel is in the form of rods of uranium oxide contained in a metal cladding. With HTRs, the moderator is in the form of a coating for the fuel and is an integral part of it, unlike the PWR where the water flows past the fuel. This gives some safety advantages as the moderator which controls the reactor cannot be separated from the fuel. This combination of helium coolant, graphite moderator and ceramic fuel allows the reactor to operate at very high temperatures, 750ºC compared to 300ºC in a PWR. This in turn means that a much higher proportion of the energy from the core can be turned into electricity (the thermal efficiency), 40 per cent compared to 34 per cent for a PWR. It also means that a much higher proportion of the U235 can be split, giving high fuel `burn-up'. This means that the reactors are more economical in their use of uranium and create a much lower volume of used, or `spent' fuel.   All high temperature reactors built to date have used highly enriched uranium (HEU) - more than 90 per cent U235. While this may lead to good uranium utilization, such material is a serious weapons proliferation risk. South Africa's nuclear bombs were built using HEU. The use of such a material as a basis for nuclear power plants to be exported round the world would raise huge concern on proliferation grounds and it is unlikely that the international community would allow South Africa to go ahead using such material. For its PBMR, Eskom plans to use 7-8 per cent enriched uranium, very different to the type of fuel used in HTRs so far.   Like most purpose-designed reactor types, but unlike the submarine-derived PWR, the PBMR would avoid the need for an annual shut-down for re-fuelling, by re-fuelling while the plant is operating, `on-line'. In theory, this should mean that extra power can be produced. In practice, on-line refuelling has not always worked out well because the machines for doing it are complex, expensive and prone to break-down. Also, the time required for maintenance, which is carried out at the same time as refuelling, usually exceeds the time required for re-fuelling so on-line refuelling would not reduce the amount of time the plant is off-line.   For example, in Britain, the Advanced Gas-Cooled Reactor (AGR) was designed to refuel on-line, at full power. But more than 20 years after the first plant went into service, the regulatory authorities still do not allow refuelling at full power because of safety concerns. Ironically, in 1965 when the AGR was chosen, it was the extra output that was expected to be produced because of on-line refuelling, that swung the economic case in favour of the AGR over US designs. This reduced the overall generation cost of the AGR by a small fraction of a penny. This experienc