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. The US Department of Energy (DOE) launched the Generation IV initiative in 2000. Today, it comprises 10 member countries plus Euratom, with the goal of developing innovative nuclear systems (reactors and fuel cycles) likely to reach technical maturity by 2030. The ten countries are: Argentina , Brazil , Canada , France , Japan , the Republic of Korea , the Republic of South Africa , Switzerland , the United Kingdom and the United States . Six nuclear systems were selected; ·        Gas-Cooled Fast Reactor ·        Lead-cooled Fast Reactor ·        Molten Salt Reactor ·        Sodium-Cooled Fast Reactor ·        Supercritical-Water-Cooled Reactor ·        Very High Temperature Reactor With the lofty goal of achieving significant improvements in economic competitiveness, nuclear safety, uranium resource economy and in reducing long-life radioactive waste. A technology roadmap (http://gif.inel.gov/roadmap/), initially prepared at the request of the US Congress, is now used as a basis to structure and share the R&D effort among the participating countries, in order to develop the selected nuclear systems. In the future, other countries or international authorities could join the Generation IV International Forum and the related R&D effort. Generation IV – pooling international R&D The founding principle of the Generation IV International Forum (GIF) is its members’ recognition of nuclear technologies’ role in satisfying the world’s increasing energy needs – in the context of sustainable energy development and climate change prevention. This principle, laid down in the GIF charter, expresses a common will to create a framework for international R&D. This framework’s role is to define, develop and enable the deployment of Generation IV nuclear systems. Most of the research involved will develop within this multilateral framework, over the coming decades. What is Generation IV? Nuclear systems can be classified according to the generation they belong to. The US DOE distinguishes four generations: ·         Generation I was operational before the 1970s and made use of natural uranium, to avoid the need for enrichment; ·         Generation II comprises light water reactors deployed since the 1970s. They are still in operation; ·         Generation III involves optimizing the current reactors, in terms of economics and safety. These reactors are likely to be deployed before 2010; ·         Generation IV comprises nuclear systems likely to reach technical maturity by 2030. Their design will take cognizance of the progress made in economics and safety. In addition, the aim is for these reactors to support sustainable energy development worldwide, and to open up the range of nuclear systems’ applications to hydrogen generation for transport (in addition to electricity production). Generation IV’s goals International consensus has been reached on the on the general goals and criteria to be met by Generation IV nuclear systems. They will have to be: ·     sustainable: the systems should offer efficiency in the use of the natural resources, and should minimize environmental impact (and, at the same time, minimize waste in terms of mass, radio toxicity, residual power, etc.); ·     economically viable: economic considerations are: the generating cost, which should be competitive when compared with other energy sources; and the capital investment cost, which should be low enough for the nuclear system under development to remain accessible to a large number of countries (total investment cost and specific investment cost ($/kWe) refers); ·     safe and reliable: it is mandatory that future reactors perform at least as well in terms of safety and reliability as current reactors. In particular, key focus is to be placed on eliminating, as far as possible, the need for public evacuations from areas outside nuclear sites – in the event of an accident, whatever its cause and extent of gravity; ·     Resistant to proliferation risks and likely to be easily protected from external attack. Besides electricity generation, the Generation IV systems will offer potential for the generation of hydrogen from water for use in transport, seawater desalination, and heat generation for industrial processes. The selection of Generation IV systems From April 2001 to October 2002, technologies likely to meet the above-mentioned criteria were identified by means of the roadmap. The six selected concepts will be developed, in multilateral co-operation, during a second phase of the Forum’s activity. The six concepts, as presented in Figure 1, are: ·         GFR : Gas-cooled Fast Reactor system cooled with helium; ·         LFR: Lead Fast Reactor cooled with lead or lead-bismuth eutectic; ·         MSR: Molten Salt Reactor fuelled with molten salts; ·         SFR: Sodium Fast Reactor; ·         SCWR: Super-Critical Water-cooled Reactor; ·         VHTR: Very High Temperature Reactor cooled with helium at 1000°C at the core outlet, for efficient production of hydrogen. Among the six selected systems, three have fast neutron spectra: the GFR , SFR and LFR. Two advanced gas-cooled systems ( GFR and VHTR) rely on R&D which, to a large extent, follows the same path. The SCWR was retained with a thermal neutron spectrum as an intermediate step and a fast neutron spectrum as an ultimate goal. The MSR stays in the running as a non-conventional system. With the exception of the VHTR, all these systems will be operated with a closed fuel cycle. Key comments relevant to the selection of the six Generation IV systems: ·     Among the retained goals and criteria, sustainability has been the most discriminating. This is demonstrated by the majority of the systems having fast neutron spectra and closed fuel cycles. The grouping of systems in families – according to homogeneity in performance and R&D needs – turned out to be important for optimizing R&D efforts and structuring recommendations along the lines of federal guidelines. Selecting two gas-cooled systems ( GFR and VHTR) was an acknowledgement of the interest in this coolant for high temperature applications. A strongpoint governing this selection is the consistency of the technology they employ, enabling a sizeable common R&D pathway to be followed. The issues which prevented the more innovative systems from being selected concerned important uncertainties on: their definition and performance, as well as on the prospects of overcoming obstacles to their viability before the 2030 realization deadline. Shared R&D Efforts Two main phases of R&D were identified: ·     The ‘feasibility phase’ dedicated to resolving technology showstoppers; The ‘performance phase’, aiming at confirming and optimizing the systems’ performance (which was the criterion emphasized in the selection process). The feasibility studies for the more mature systems – i.e. the Sodium Fast Reactor (SFR) and the Very High Temperature Reactor (VHTR) – should last until 2008 and 2012 respectively; and the performance studies, until 2015. Regarding the more futuristic systems – i.e. those having fast neutrons and gas coolant ( GFR ), lead (LFR) and supercritical water (SCWR) – the feasibility studies will continue through 2013 and 2015, and the studies of optimization until 2020. The timeline for the molten salt system will be longer. While nuclear R&D organizations will play an essential role in both the feasibility and the performance studies, there will be substantial participation from universities, other research organizations and manufacturers. Countries that are not currently GIF members will be able to join the R&D effort at this stage. Following the feasibility and performance phases, it is expected that international consortiums of manufacturers and R&D laboratories will support demonstrations of the key technologies. They will also identify which Generation IV systems they are interested in launching commercially. Progress in preparing for the next phase of GIF The GIF Policy Group meeting, held in Zurich held on January 27 -27, 2004, led to progress on three main topics: the co-operation agreement at system level, the governance of the Forum, and relations with other organizations. In Zurich , the US DOE presented a draft system agreement – jointly prepared by its State and Trade Departments – covering the first 10 years of co-operation (and providing for extensions in five-year increments). This project of agreement will establish the R&D framework required to address Generation IV system feasibility issues, as well as to confirm system performance, established during the system selection process. Future phases of demonstrating and commercializing the six selected nuclear systems will be the subject of further agreements. The parties to this system agreement are intended to be governmental entities or mandated national laboratories. GIF members are to be invited to give their input on the draft system agreement, which is expected to be finalized by mid-2004 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. Details regarding the PBMR design can be found on http://www.pbmr.com/. 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 experience will not necessarily be repeated in South Africa but it does demonstrate that refuelling on-line can be a difficult process and that any projected economic advantages to on-line refuelling should be treated with some remaining sceptical.  The Track Record of High Temperature Reactors In nuclear power, as with any other field of technology, design concepts that look good on paper cannot necessarily be turned into viable and economic technologies. It is therefore important to examine attempts by other countries to turn this apparently attractive concept into a commercial technology. The clear intrinsic advantages of the HTR, namely (a) high thermal efficiency, (b) economical use of uranium and (c) better safety, have meant that from the earliest days of civil nuclear power, this class of reactors has been examined carefully by almost every nation that has tried to design nuclear power plants. The first prototype plants of this type were ordered in the late 1950s. The USA and Germany have gone as far as building prototype plants of a commercial size, about 300 MW (a third the size of each Koeberg unit and three times the size of the proposed South African PBMR). German experience is particularly relevant to South Africa because it is German technology which has been sold to South Africa and forms the basis of the PBMR. The UK and Japan have built small-scale prototype reactors for research purposes which do not produce electricity. France seriously considered developing its own commercial scale design of HTR in the late 1960s as an alternative to importing PWR technology. Of the countries which can claim to have nuclear design capability, only Russia and Canada have shown little or no interest in the HTR. Today, the USA, Germany, the UK and France have now abandoned all interest in HTRs, while Japan's development programme is very slow and there are no plans to build commercial power plants. The USA: The USA was the first country to build a HTR power plant, the Peach Bottom 1 plant, ordered in 1958 and completed in 1967, which produced about 40 MW of electricity. Like all plants of this design in the USA, it was built by General Atomic (a company owned by Gulf Oil) and operated until 1974. The operating record of the plant seems to have been fairly good and the plant has now been completely decommissioned. None of the US plants is of the pebble bed design. Confidence in nuclear technology of all types was then so high that even before this plant had been completed, a successor, about 8 times as large was ordered. Fort St Vrain was ordered in 1965 and designed to produce 330 MW. It was owned by a utility, Public Service of Oklahoma but about half the construction cost was paid by the US government. It went critical in January 1974, but did not generate its first power until December 1976 and was only declared commercial (handed over from the supplier to the owner) in 1979, a good indication that all was not going to plan. For a commercial nuclear power plant, the time from first criticality to commercial operation should be less than 6 months (it was four months at both Koeberg units). However, confidence in nuclear technology was undiminished and at the time, the USA was undergoing a huge surge of nuclear orders. In the peak year for orders, 1974, 41 units were ordered. Ironically, only 9 of these plants were completed and all subsequent orders in the USA (a further 41 plants) were cancelled. The plants were cancelled because the costs were too high or electricity demand was not sufficient to justify them. Orders for full-size plants of the HTR design, without any government subsidy, were first placed in 1971 and by 1974, eight orders had been placed, four for units of 770 MW and four for units of 1160 MW. Little or no progress on these plants was made and with problems at Fort St Vrain becoming apparent, all were cancelled in 1974-75. Fort St Vrain continued in service from 1976 until August 1989 when its high costs and appalling reliability finally persuaded the owner to give up the struggle and retire the plant, which has now been largely decommissioned. Over its 10 years of commercial service, its average load factor (power produced as a percentage the power the plant would have produced had it operated uninterrupted at full power) was 15 per cent. Typically a plant owner would expect a load factor of about 80 per cent from a nuclear power plant. There was no single overwhelming factor that led to its failure, more a series of different equipment problems. Despite this bad experience, in 1991, when the US government decided it needed to put money into new reactor development, it looked at three or four technologies, one of which was the Gas Turbine Modular High Temperature Reactor (GT-MHTR). The design was close to the PBMR because it used a gas turbine rather than a steam turbine and was planned in modules, but used fuel rods rather than pellets. This would have been developed partly to consume plutonium taken from dismantled bombs and partly as a civil reactor. The technology was developed until 1995, although it was close to losing funding on several occasions, and in August 1995, the US government finally withdrew support. It used the few resources it was prepared to spend on nuclear technology to support advanced PWRs and BWRs (Boiling Water Reactors, a close relative of the PWR). At the time, a National Academy of Sciences review revealed that HTR technology had received US$ 900m of government money over 30 years. It claimed that the GT-MHTR would take a long time to get a safety licence. It identified fuel as a particular problem because of the lack of any fuel production facilities. New fuel facilities would have to be licensed and built adding to the delay and cost. Germany: Germany also has a long history of HTR development dating back to the ordering of the Jülich plant, at the government research centre there, in 1959. This 15 MW plant, financed by the government, was ordered from a group led by Brown Boveri and Krupp and went critical in 1966, generating electricity a year later and continuing in service until 1989. Its reliability seems to have been good for a prototype and in 1970, its successor, sometimes known as THTR-300, Uentrop or Schmehausen was ordered. This too was subsidised by the government but also involved utility funding. The industrial grouping behind it, HRB, again centred on Brown Boveri but with General Atomic support. Subsequently Siemens produced modular designs involving pebble bed reactors but none were built. THTR-300 went critical in September 1983, but was not connected to the electricity grid until November 1985 and was only declared commercial in June 1987. From June until October of that year, it operated at about two thirds full power, suffering a range of problems including difficulties with the fuel circulation system. It restarted in January 1988 for a couple of months, again running at about two thirds of its full power rating, until more repairs were necessary to the fuel circulation and collection system. It ran for another five months and was shut down due to damage in the gas ducts. Repairs were completed by February 1989. But the plant remained closed on the orders of the safety regulator because of concerns about safety and the unwillingness of the various owners of the plant, including the federal government, to continue to provide subsidies to operate the plant. In 1990, the plant was permanently closed and is being decommissioned. Siemens and ABB (the new name for Brown Boveri) pooled their expertise on HTRs to form a new company called HTR Gmbh. Their strategy appears to have been to license the technology to countries such as the then Soviet Union, China, Japan and South Africa. The UK: The UK was a pioneer of nuclear technology. Its first nuclear power plants were scaled-up versions of the plants built to make plutonium for bombs. This used graphite as the moderator and carbon dioxide gas as the coolant. Nine power stations were built using this technology, but the technology was only seen as a stop-gap. Three new technologies were developed to working prototype scale, including the Dragon HTR. This was ordered in 1957 and completed in 1964. It was a research reactor with no electricity generation facilities and ran until 1974. Anecdotally, it was known as a plant that leaked radiation and another design was chosen in 1964 to form the basis of the civil nuclear power programme in Britain. Since then, HTRs have not been seriously considered in Britain. France: France followed a very similar route to Britain, developing its first civil nuclear power plants from plutonium producing reactors. Like Britain, it too had to choose a new technology route by the mid to late 1960s. The French nuclear research establishment strongly favoured HTRs, but strongly influenced by the utility, American PWR technology was chosen and, as in Britain, HTR technology was abandoned Japan: Japan has persisted with a wide range of nuclear technologies for much longer than other countries. It imported British technology for one commercial plant in the 1960s, but since then, all commercial orders have been for US designs, PWRs and BWRs. Nevertheless, it has built a medium size plant of its own design (165 MW) using heavy water as moderator. This was completed in 1979 and for many years there was talk about building a plant of 600 MW of this design. This technology line has now been abandoned. A prototype fast reactor, Monju (280 MW), was completed in 1995, but an incident at the plant in December of that year drained public and regulatory confidence in the plant and it is highly unlikely the plant will run again. A third line of reactor development using HTRs of a Japanese design has been underway at a slow pace since about 1990. A prototype reactor producing about 30 MW thermal power but no electricity was completed in 1998, some 3 years later than scheduled. China: For more than 20 years, China has had ambitious plans to launch a programme of civil nuclear power plants and from 1980 onwards, forecasted that about 20 nuclear power plants would be in service within 10-15 years in China. There is still little to show for their efforts. Two imported power plants were completed in 1993-94 (the same design and supplier as Koeberg) and one plant of a Chinese design was completed in 1992. The potential size of the Chinese market and the dearth of nuclear orders in the West mean that nuclear vendors continue to pursue orders in China despite the political, economic and commercial problems that arise. In 1989, China signed a licensing deal with HTR Gmbh to develop HTRs in China. There is little to show for these efforts yet. Development of Nuclear Technologies The history of nuclear power development has been one of unfulfilled promises and unexpected technical difficulties. The ringing promise from 1955, of `power too cheap to meter' is one that has come back to haunt the nuclear industry. With most successful new technologies, people confidently expect that successive designs become cheaper and offer better performance. This has not been the experience with nuclear power: costs have consistently gone up in real terms and processes which were expected to prove easy to master continue to throw up technical difficulties. The issues surrounding waste processing and disposal which at first were assumed to be easily dealt with, remain neglected. Despite this history of unfulfilled expectations, two factors have meant that nuclear power continues to be discussed as a major potential energy source. First, the promise of unlimited power independent of natural resource limitations and second, the attraction to engineers and scientists of meeting the technological challenges that are posed. However, in the developed world, patience with nuclear technology is running out. Governments are no longer willing to invest more tax-payers' money in a technology which has provided such a poor rate of return. Electric utilities cannot simply pass on development costs to consumers. Equipment supply companies, which have generally made little or no money from nuclear technology, are unwilling to risk more money on developing technologies which might not work well and which might not have a market. There is still talk about new nuclear technologies, but a critical look at the real resources going into them shows that little money is now being spent. Other Technological Aspects In this first section, the track record of the HTR has been examined and it is clear from this that the world's leading nuclear countries have all examined HTR technology in some depth, especially Germany and USA, arguably the two leading nuclear nations, and none has been able to make a success of it. It is not impossible that South Africa could succeed where so many others have failed, but it seems inappropriate that public money should be gambled on such a risky technology. However, the technological risk does not end with the reactor. No facilities exist to manufacture the nuclear fuel and these would have to be set up in South Africa. The German reactor of this basic design experienced a number of fuel problems in its short life, so it cannot be assumed that manufacturing fuel pellets will be simple. Even the conventional part of the plant, the gas turbine, would be a new product developed at Eskom's expense. Eskom's publicity describes this part of the plant as using the `standard Brayton cycle' implying a well-proven standard product. No gas turbine using helium has ever been operated and a number of its features are substantially novel. Eskom did request the major manufacturers to tender for a full product with guarantees but it appears that none of them responded. One supplier suggested that research, funded by Eskom would be needed before a commercial product could be designed and produced. The Economics of Nuclear Power The economics of nuclear power is a highly contentious area. It is often difficult to establish independently verified estimates of the basic construction costs and the operating cost. In addition, the results are crucially dependent on the accounting and investment appraisal assumptions such as the rate of return on capital that is sought (the discount rate) and the life-time of the plant. These latter factors are of particular relevance to nuclear power because the main element in the cost for each unit of electricity generated is that associated with building the plant, the capital cost. The shorter the expected life-time and the higher the discount rate, the higher these fixed costs will be. In a monopoly system, the assumed life of the plant can be the expected physical life-time because there will be nothing to stop the owner running the plant until it is worn out. In a competitive system, the plant may have to be retired much earlier if it cannot compete with new plants. The running costs of nuclear power plants are difficult to establish because most electric utilities regard this data as commercially confidential. However, in the USA, utilities are required to publish fully authenticated running costs. In 1997, the cheapest to run nuclear plants cost about 1c/kWh (0.6p/kWh), while the average was about 2.4c/kWh (1.5p/kWh). Of this, about 0.4-0.6c/kWh was fuel cost while the rest, 0.5-1.8c/kWh, represented the non-fuel cost of operation and maintenance (wages, spare parts etc.) Government owned utilities have usually been able to invest money at very low rates of return on capital partly because new power stations were seen as a safe investment and partly because, for a variety of reasons, governments have tended to require a lower rate of return on capital than private industry. Thus, in Britain before privatisation, the national utility, the CEGB, could invest at a 5 per cent real (net of inflation) rate of return and recover the costs over 35 years. After privatisation, it is known that private investors are looking for about 12-15 per cent real return and recover the capital over 15-20 years. A simplified scheme can be used to estimate the fixed cost of electricity from nuclear power stations. We can assume that the capital is repaid in equal annual payments over the life-time of the plant. For the interest payments, we can assume that the average amount owed over the life-time of the plant is half the total construction cost. If we do some simple arithmetic based on the cost of Sizewell B, the consequences of the change in lifetime and discount rate are clear. Each kilowatt of capacity at Sizewell cost about £3000 to build and will generate about 6000 kilowatt hour (kWh) per year. If we recover the costs over 35 years and charge 5 per cent interest, the cost in pence per kWh simply to repay fixed costs and taking no account of running costs, will be: (Interest paid based on the average amount owed + capital repayment) / units of output per year = fixed cost per kWh (1500 x 100 x 0.05 + 3000 x 100 / 35) / 6000 = 2.7p/kWh During the process of getting public approval for Sizewell B, the government, realising that its discount rate was well below commercial rates, raised the level to 8 per cent. This change alone raised the fixed cost to 3.4 pence. If we do the same calculation with an interest rate of 12 per cent and recover the cost over 20 years, generous assumptions in a competitive market, the cost per kWh is 5.5p/kWh. With a 15 per cent discount rate and a 15 year life, the fixed cost is 7.1p/kWh To put these figures in context, the total cost (fixed and running) of a new coal plant when Sizewell B was first planned was about 3.5p/kWh (British coal was then about four times as expensive as South African coal). So, if the running costs of nuclear were as low as the best US plants, using the original assumptions (5 per cent discount), Sizewell B might have been economic. By the time of privatisation, new gas-fired plants were being bought and these were expected to generate at about 2.9p/kWh and so, with an 8 per cent discount rate, the total cost of power from Sizewell B was perhaps 50 per cent more expensive than gas-fired generation. By 1996, the cost of gas-fired plants and of gas had come down and their efficiency had gone up such that the total generation cost was now about 2.2p/kWh, a quarter of the cost of nuclear power using the same assumptions on life-time and discount rate. The importance of operating performance should also be clear from these examples. If instead of 6000 kWh per year, the plant had only produced 3000 kWh, the fixed costs would double. Over its life, Fort St Vrain averaged about 1300 kWh per year. It can easily be seen that nuclear power is so far from being economic in Britain, it is not a serious option for any utility. In France where large numbers of nuclear power plants have been built, construction costs appear to be much lower (they are not independently authenticated). If plants could be built for half the cost of the British plant and generate 7500 kWh per year, the cost per kWh would still be 75 per cent higher than gas-fired plant. So even in the most successful nuclear countries, nuclear power appears to be uneconomic in a competitive market. The key economic assumptions that have gone into Eskom's estimate for the PBMR are, (a) the construction cost is assumed to be about US$1000 (£625) per kW, (b) the plant life is 40 years, (c) the discount rate is 6 per cent and (d) the assumed availability is 95 per cent (8300 kWh per year). The expected running cost is not fully documented, only the fuel cost which is estimated to be about 0.4c/kWh, equal to the cheapest US nuclear power plants, is included. The total running cost is therefore likely to be about 1c/kWh (0.6p/kWh). For comparison, this means Eskom expects the PBMR to be built for about 20% of the cost of the most recent British nuclear power plant and they expect it to be able to achieve a reliability better than any nuclear plant in the world has ever achieved over several years. At £1=$1.6, this gives a fixed cost, using these assumptions, of about 0.4p/kWh. If we accept these remarkable construction costs and availability, but put in commercial discount rates and life-times, but at the low end of the likely values, 12 per cent and 20 years, the fixed cost doubles to 0.82p/kWh. If we use the values for discount rate and plant life-time generally used in Britain now, 15 per cent and 15 years, the fixed cost increases to 1.1p/kWh. Simply by changing the investment appraisal parameters to ones more appropriate, much of the cost advantage of the PBMR over CCGTs has largely disappeared. The importance of the life-time is clear, but the discount rate may be seen as a rather esoteric debate which it is hard to relate to. However, the reality is that the choice of discount rate is at the heart of the debate about how national resources are allocated. The amount of investment capital available to a country is not unlimited. If money is spent on low-return projects, money will not be available to higher return projects and the economic growth of the country will suffer. The discount rate is as high as it is in Britain because that is the rate of return that the projects can achieve. If the government (and Eskom is owned by the South African government) spends money on low-return projects, there could be two effects: first, money will not be available to the private sector to invest in projects that will generate more wealth; and second, public sector projects, perhaps even within Eskom, such as urban and rural electrification, with a much better rate of return will not be funded. It is not clear how fully the PBMR has been costed and whether equipment suppliers have been identified. However, even if suppliers are known and costs have been quoted, all the history of nuclear power suggests that these costs will not be an accurate reflection of the actual costs. Two main factors, uncertainty about the features that the safety regulator will demand and the risk that, with an unproven design, unforeseen difficulties will arise, mean that no credible supplier would quote a guaranteed fixed cost. Even if such guarantees were given, there must be some doubt about whether they were worth the paper they were printed on. Even a small nuclear power plant such as the PBMR would produce electrical output worth about £20m per year. Eskom plans these plants in clusters of ten so any design fault would probably be repeated ten times over before it was discovered. If this resulted in a delay of only a year to construction, the value of the lost power would be £200m which the supplier would be liable for. Few companies have the resources to back such a guarantee and even fewer would choose to do so. The HTR has undeniable intrinsic safety advantages which probably make a catastrophic accident such as occurred at Chernobyl impossible. However, these intrinsic safety advantages are not sufficient to guarantee the safety of the plant. A competent safety regulator would not be prepared to give approval for the design until the full detailed design was available and the plant could not get an operating licence until it was built. There is ample experience in the West of plants of similar basic design to those already in operation, running into construction cost and time overruns because detailed design points were not acceptable. The German experience with the THTR-300 plant, the fore-runner of the PBMR which had the same intrinsic safety features is relevant here. This plant was licensed and in service for a year when problems at the plant led to the withdrawal of the operating licence, a factor instrumental in its closure soon after. The British experience with the AGR is particularly salutary in this respect. When the Dungeness B plant was ordered in 1965, a prototype plant of this design was operating, apparently successfully. The plant was ordered under fixed cost terms from a British supplier. The detailed design proved to contain serious errors which resulted in constant redesigns throughout the construction period. The supplier and two successor companies went bankrupt, so cost guarantees proved worthless. The plant was finally declared commercial in 1988 after 23 years of continuous construction and huge cost overruns, all of which were paid for by electricity consumers. The lengthy construction period (some of the equipment was obsolete before the plant entered service) and the numerous design errors mean that the plant will never operate as designed and in 1998, one of its better years, the load factor was only 42 per cent. The reliability levels projected by Eskom are also hard to justify based on Eskom's track record with the Koeberg plant. In 1996, the latest year for which there is full data, the average load factor for the world's nuclear power plants was 77 per cent. Over the 12 years that Koeberg had been in service, the plants averaged a load factor of 58 per cent. In 1997 and 1998, the plants did rather better, but neither was in the world's top 50 plants. There is therefore nothing in Eskom's record to suggest that it is capable of world-beating performance with nuclear power plants, especially with a new and unproven design. If we assume that Eskom's construction cost estimate is half what costs would really be - this would still make the PBMR the cheapest nuclear plant in the world to build - and we assume the load factor achieved is a little above the average of plants in the rest of the world (7000 kWh per kW per year) and we recalculate the fixed costs, the equation is as follows, using a 12 per cent discount rate and a 20 year life-time 625 x 100 x 0.12 + 1250 x 100 / 20 / 7000 = 2.0p/kWh or, using a higher discount rate (15 per cent) and shorter life-time (15 years), 625 x 100 x 0.15 + 1250 x 100 / 15 / 7000 = 2.5p/kWh We can compare this with the full cost new gas-fired plant in Britain of about 2.2p/kWh. It is clear that even if South Africa could build plants at less than half the cost of Britain, if it could operate them at above the world average level of reliability, and if running costs were as low as the best US plants, gas-fired plants would be much cheaper. The World Market for Nuclear Power Plants Eskom's evaluation of the PBMR is based on projections of an annual market of 30 units, 10 for installation in South Africa and 20 in the rest of the world. It is therefore important to establish what the world market for nuclear power plants is and what share South Africa might hope to gain from it. If we start with Europe, 10 countries have built nuclear power plants. Austria closed its plant without operating it after a referendum. Italy closed its three plants after a referendum. Sweden is committed to closing its plant early after a referendum. The newly elected German government has committed itself to phasing out nuclear power. The Netherlands and Switzerland are also likely to phase out nuclear power, while the Spanish government ordered the abandonment of work on several unfinished plants in the 1980s. As argued above, new nuclear orders in Britain are not feasible, leaving only Finland and France as the only countries where new orders are possible, although not inevitable. France has spent huge amounts of money developing its own nuclear capability and it is inconceivable that, if orders were placed, it would not use French companies. For more than 20 years, Turkey has talked about placing nuclear orders and frequently, deals are said to have been imminent. So far, these have all come to nothing and it seems unlikely that Turkey will be a major market for nuclear power in the next decade. In North America, no orders not subsequently cancelled have been placed since 1974. Canada has developed its own technology which is now running into severe problems on the economics and safety side with several units shut down for several years as a result. It is barely conceivable that any new orders would be placed. In the USA, more than 100 nuclear orders were cancelled, losing consumers billions of dollars. As in Canada, the electricity industry is being liberalised and many existing nuclear plants are being categorized as stranded assets. The two Mexican units took more than 20 years to build and cost over-runs were huge. Given this poor record, new orders for nuclear power in any of these countries are not feasible. In South America, Brazil and Argentina have built nuclear power plants. Argentina has two operating plants and has been struggling to finance completion of a third plant, of Canadian design for more than 20 years. Brazil has one operating nuclear plant which, over a 20 year life, has an average availability of about 20 per cent. It may complete a second plant of German design which started construction in 1975 and will cost about US$9bn, making it about the most expensive nuclear plant built. These countries are unlikely to want to repeat their sad experience with nuclear power, nor are their neighbous likely to launch new programs. In Africa, only South Africa is actively pursuing nuclear power and the chances of nuclear sales outside South Africa are minimal. This leaves only Asia as a possible market for nuclear power. The two giants of the continent are India and China, both with nuclear power programmes. India and Pakistan both acquired nuclear power plants in the 1960s but after India exploded a nuclear bomb in 1975, all international nuclear contacts were cut. As a result it has tried to develop its own designs based on the plant it bought from Canada. It now has about 10 small (200 MW) plants in service. All have seriously overrun their construction time and cost forecasts and have been hopelessly unreliable. India is now trying to buy a plant from Russia, but it is unlikely that either side has the cash to carry out this project. Pakistan has recently bought a small plant from China of Chinese design. Like India, its poor record on nuclear proliferation makes it largely impossible for Western countries to do business there with nuclear technology. China has, for the past 20 years, had ambitious plans to build nuclear power plants of imported design and of its own design. These have resulted in few orders so far: two plants are in service of French design, two more French plants are on order and two Canadian plants are on order. One plant of Chinese design, a 300 MW PWR, is in service, but is currently off-line with serious equipment problems. One plant of this design was sold to Pakistan and China is planning to build further units of this basic design, but double the size. All nuclear vendors are active in China because of the potential size of the market, but it is doubtful whether China can finance a significant nuclear power program. As noted previously, Japan has developed a number of its own nuclear technologies, but none of these has been ordered for commercial use. All its operating plants are of US design and Japanese companies such as Mitsubishi, Hitachi and Toshiba have spent large sums of money over the past 30 years developing an understanding of these technologies as well as manufacturing facilities for them. While Japan now has a large number of operating plants (53 at the beginning of 1999), public opposition and problems in finding sites due to seismic activity mean that further orders are now very difficult. There is no room on established sites for further plants and, now, only two plants are under construction. If Japan does order further plants, they will almost certainly be more units of US design or units using a new Japanese design. Of the other Asian countries, South Korea and Taiwan have nuclear power plants in service. Korea has 14 plants in service and another 3 under construction. It has expended a large amount of effort transferring US technology in and has built up full manufacture facilities. It is highly unlikely that future nuclear orders would not be supplied using these facilities. Taiwan has six plants in service and two on order. When these two plants are complete, there will be little scope for further nuclear plants. Other Asian countries, such as Thailand and Indonesia have, for 20 years or more, discussed the possibility of ordering nuclear plants. However, there is little to suggest that these discussions will soon be turned into nuclear orders. The Market for South African Nuclear Power Plants It seems likely that the world market for nuclear power plants may be no more than one or two units a year. It is not clear whether South African designed plant could be expected to win any of this market mainly because of the conservatism of the market. The accidents at Three Mile Island (USA) and Chernobyl (Ukraine) have alerted nuclear buyers to the economic risk arising from such accidents. Following any serious accident, all plants throughout the world have to demonstrate (if that is possible) that they are not vulnerable to such a set of events. This can be expensive and time-consuming. If modifications are required, there is some comfort in owning a type of plant widely installed elsewhere whose owners will pool resources to solve the problem quickly and efficiently. The record of rivals to the established designs, the PWR and the BWR, is poor especially for the HTR and the breeder reactor, designs with many theoretical attractions but which do not seem able to be translated into a working commercial design. Buyers therefore have a strong incentive to stick with tried and tested designs. Buying a new design from a country with no track record in nuclear reactor technology appears an enormous risk. Waste Disposal When nuclear power plants were first planned and built, there was little consideration of how waste would be dealt with and worn-out plants removed. It was assumed that new technologies would emerge and costs would be small. In most countries, waste is divided into three categories. Low-level waste (LLW) is not strongly radioactive and humans would require significant exposure to suffer any health consequences. After a few decades, the radioactivity has generally decayed sufficiently that the material presents little hazard. Intermediate level waste (ILW) is much more strongly radioactive, it remains radioactive for much longer and must be dealt with much more carefully. High level waste (HLW) is not only strongly radioactive but it also generates large quantities of heat. While activity does decay to some extent, HLW must be kept away from human contact indefinitely. Most countries have had some limited means of dealing with LLW for several decades. Medical and scientific uses result in small quantities of LLW, the isotopes themselves, but also everything they come into contact with, such as gloves and lab coats. At first, this material was simply bull-dozed into holes in the ground and covered. Now, greater care is taken and it is placed in sealed concrete containers and usually buried in shallow ground. It is assumed that by the time the concrete containers have failed, the radioactivity is no longer a hazard. These original dumps are now becoming full: their capacity can be eked out by compaction techniques, but most countries are now searching for new sites. This is invariably politically contentious and few countries have had any success in the last couple of decades in siting new dumps. In Britain, it was decided in the mid-80s that all LLW would be disposed of in a new deep engineered facility, which would also take all ILW, when the existing facility at Drigg was full. This would clearly raise the costs by a large amount, probably an order of magnitude. However, proving that the geology of such a facility would be stable over a long enough period that it could be assumed there would be no risk that radioactive material would get into the ground water, is a difficult task. It was planned that a test hole be drilled and the geology observed over a decade before the facility was built. A public inquiry rejected the case in 1997 for the one site selected in Britain. There is now no investigation for alternative sites. If the process started tomorrow, an optimistic time-table might require 5 years to identify another potential site, a couple of years for public consultations (the siting would be bitterly resisted), 15 years to build and observe a test drilling, 5 years to build a commercial facility. Britain therefore cannot have a new LLW facility until 2025, by which time LLW will be piling up in temporary stores. As the standards for LLW disposal have been raised, the costs have gone up. In the last 10-15 years, LLW disposal costs in the USA have been rising at about 6-7 per cent per year in real terms, that is, doubling every 10 years. There is little sign that this price escalation is falling away and, while waste disposal is still quite a small part of nuclear generation costs, if this process is not checked, it could become significant. ILW is typically material that has been in close contact nuclear fuel, for example, steel vessels. There are no facilities for final ILW disposal in Britain or in most other countries - the only modern facility is a deep repository in Sweden. The material is presently stored in temporary containers on the surface awaiting the construction of the facility described above. Most such material was temporarily packed in containers designed to last a decade or two. The late completion of the disposal facility will mean that this material will have to be unpacked and re-packed at significant expense and will be a hazard over that period. HLW represents the most intractable technical problem, although the volumes of material are much lower than for the other categories. Essentially, HLW is either spent fuel or the product of the reprocessing of spent fuel. Disposal facilities must be designed such that for thousands of years, there can be no risk that the material can get out of its containers and get into the ground water where it would come into contact with humans. There is a difficult philosophical debate about whether the material should be retrievable or not. If the material is retrievable, if anything goes wrong with the storage facility, it can be retrieved and made safe, but the material is accessible and can be misdirected. If the material is not retrievable, the pros and cons are reversed. There is no clear winner to this debate yet. At present, no country in the world has identified a site for the disposal of HLW and all material is stored in temporary surface facilities. The technical rationale for this is that the spent fuel is still generating too much heat for it to be disposed of - any containers would come under intense strain because of this heat and would not be able to last the thousands of years required. Thus, in Britain, a decision was taken in about 1980 not to even look for sites for 50 years. However, until sites are identified, the geology proven and the methods of containment subjected to proper public scrutiny, the costs cannot be predicted with any confidence, nor can it even be certain that the process will be politically feasible. Of particular relevance to the waste debate is the process of decommissioning plants at the end of their life and removing all radioactive material for disposal in proper waste disposal sites so that the land can eventually be released for unrestricted use ('green-field' status). Until this has been done, there is a risk that radioactivity from the plant will leak into the environment damaging the ecology. Decommissioning does generate large quantities of LLW and some ILW. There is almost no experience in the world of decommissioning a commercial scale plant that has operated over a full life-time to green-field status. As with waste disposal, estimated costs are escalating rapidly. If the costs are accounted for properly from the beginning of operation of the plant, they do not have a large impact on the economics of nuclear power. Under the `polluter pays' principle, this can only be done by setting up a `segregated' fund (one that cannot be drawn upon by the plant owner for other purposes) and placing the funds in low risk investments so that when decommissioning is required, there is little risk that the funds will have been lost or used for another purpose. A possible source of confusion with the spent nuclear fuel is the role of reprocessing. The rationale for reprocessing was mainly that it separated out from the spent fuel plutonium, which could be used to make bombs, or used in fast reactors. It does not destroy radioactivity, it merely separates out the fuel into its constituent parts, some of which might have a use, e.g. plutonium, but most of which still has to be disposed of as HLW. Given that weapons production from civil nuclear power plants is not politically acceptable and that fast reactors have now been abandoned, all the material still has to be disposed of. Reprocessing creates large quantities of LLW as all the material involved in reprocessing becomes LLW. It is a very expensive process which has occasionally resulted in leakage of radioactivity into the environment. Most countries now acknowledge that the cheapest and safest way of dealing with spent fuel is to dispose of it as HLW without any processing. Overall, the political, technical and economic feasibility of disposal of all types of waste and of decommissioning plants has yet to be proven anywhere in the world. A responsible policy would appear to be to carry out investigations into these processes so that there is confidence that when these processes are required, they are technically proven and the resources to carry them out are available.