Commissariat à l énergie atomique. e-den. A monograph of the Nuclear Energy Directorate

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1 Commissariat à l énergie atomique e-den A monograph of the Nuclear Energy Directorate Nuclear energy of the future: what research for which objectives? Éditions techniques

2 DEN monographs A monograph of the Nuclear Energy Directorate Commissariat à l énergie atomique, 31-33, rue de la Fédération Paris Cedex 15 Tél. : Scientific comitee Michel Alexandre, Michel Beauvy, Georges Berthoud, Mireille Defranceschi, Gérard Ducros, Yannick Guérin, Yves Limoge, Charles Madic, Gérard Santarini, Jean-Marie Seiler, Pierre Sollogoub, Étienne Vernaz, Research Directors. The following people participated in this work: Fanny Bazile, Patrice Bernard, Bernard Bonin, Jacques Bouchard, Jean-Claude Bouchter, Bernard Boullis, Franck Carré, Jean Cazalet, Alain Marvy, Valérie Moulin, Emmanuel Touron, Yves Terrien. Publishing Supervisor: Philippe Pradel. Editorial Board: Bernard Bonin (Managing Editor), Bernard Bouquin, Martine Dozol, Michel Jorda, Jean-Pierre Moncouyoux, Alain Vallée. Administrator: Fanny Bazile. Editor: Jean-François Parisot. Graphic concept: Pierre Finot. Cover illustration: Véronique Frouard. Correspondence: all correspondence can be addressed to the Editor or to CEA / DEN Direction scientifique, CEA Saclay Gif-sur-Yvette Cedex. Tél. : CEA Saclay and Groupe Moniteur (Éditions du Moniteur), Paris, 2006 The information contained in this document can be freely reproduced, with the agreement of the Editorial Board and due mention of its origin.

3 Preface After a dazzling start in the 1950s as a promising, inexhaustible, cost-effective energy source, nuclear energy was rejected by majority opinion in several countries in North America and Western Europe three to four decades later, suddenly bringing its development to a halt. Although the 1973 and 1979 oil crises marked the beginning of massive construction programmes in the countries most heavily penalized by oil imports, France and Japan in particular, they were paradoxically followed by a gap in nuclear spending, first in the United States and then in Western Europe. However, more recent oil market tensions and emerging concerns over non-renewable natural resources should have increased such spending. There are surely many reasons for this pause, which can in part be explained by the accidents in Three Mile Island in 1979 and Chernobyl in 1986, which deeply impacted public opinion. On top of this, ecological movements and Green parties made their (highly publicized) fight against nuclear energy a key part of their platform. In France, whose population, with the exception of one case, had never disputed nuclear plant construction, negative attitudes began to surface in the late 1980s concerning the nuclear waste issue. Given Andra s growing difficulties in finding an underground laboratory site, the Government decided to suspend work in favour of a one-year moratorium and submitted the issue to the OPECST (French parliamentary evaluation office for scientific and technological choices). The Act of 30 December 1991 on nuclear waste management implemented the essence of the OPECST s recommendations, in particular its definition of a diversified research programme and the basis for democratic discussion, thus helping calm the debate. That said, although it is now an accepted fact that long-term nuclear waste management is a necessity, there is still no guarantee that France will continue its electronuclear programme: for this reason, the recent energy act of 13 July 2005 merely aimed to keep nuclear options open through However, this century should be marked by renewed collective awareness that our generation s energy needs cannot be met without concern for the environment and without preserving future generations rights to satisfy these same needs. This concept of sustainable development is an inevitable challenge to our society. Today, it goes unquestioned that global warming due to increasing greenhouse gas emissions is a human-caused problem. The only remaining debate concerns the consequences of this climate change. Industrialized nations, which are for the most part responsible for the current situation, should feel particularly obliged to voluntarily take steps towards reducing emissions of these gases. Nuclear energy should gain considerable ground since, by nature, it does not produce this type of emissions and yet is an abundant, reliable and cost-effective energy source. The situation varies from country to country. On one hand, European countries such as Germany and Belgium have chosen to progressively stop using nuclear energy, even without making plans for reversibility. On the other hand, countries like China, South Korea, or, Nuclear energy of the future: what research for which objectives? 5

4 closer to home, Finland, are making huge investments in developing this technology. Furthermore, according to a recent statement by President Bush, the United States has decided to launch new nuclear power plant construction projects over the next ten years, picking up a process that had been on hold for over a quarter-century. Following France s national energy debate that took place in the first half of 2003, the parliamentary bill on energy adopted in June 2005 established the decision to build a demonstrator EPR in preparation for the switchover when currently operating plants will be shut down. Several signs lead us to believe that there could soon be a nuclear energy "renaissance", especially if the barrel of crude stays at or above the 70 USD mark. Nevertheless, the future of nuclear energy in our country, as in many others, will depend largely on its capacity to properly address the following two concerns: - First, its social acceptability: nuclear energy must be deployed under stringent safety and security conditions, generating as little final waste as possible, with perfect control of the waste that is produced in terms of its possible impact on human health and the environment. - Secondly, the availability of nuclear resources: it is important to guarantee a long-term supply of fuel, by preparing to resort to more economical natural fissile material systems which are less dependent on market fluctuations. These topics are a key part of the CEA Nuclear Energy Division s work. Indeed, this division is a major player in the research that aims to support the nuclear industry s efforts to improve reactor safety and competitiveness, providing the Public Authorities with the elements necessary for making decisions on the long-term management of nuclear waste, and, finally, developing the nuclear systems of the future, essentially fast neutron reactors, which offer highly promising innovations in waste management and raw material use. As a fervent partisan of sharing as much scientific and technical knowledge as possible to a broad public, I believe that this research work, which calls upon a diverse array of scientific disciplines often at top worldwide level, should be presented and explained in priority to anyone who would like to form their own opinion on nuclear energy. For this reason, it is with great satisfaction that I welcome the publication of these DEN monographs. Through close reading of these works, they can become an invaluable source of information for the, I hope, many readers. I would like to thank all the researchers and engineers who, by contributing to this project, helped share their experience and knowledge. Bernard BIGOT High Commissioner for Atomic Energy

5 Introduction Today energy problems are global problems. It is on the international scale that we share resources and risks, in particular those linked to climate changes caused by greenhouse gas emissions. For this reason any new generation* of nuclear energy production must be thought out based on serious projections on the international scale. Recent studies carried out by the World Energy Council or by the International Energy Agency of the OECD provide us with the following trends: An energy demand which will increase by 50 to 60% before 2020; A demand which will increase predominately in developing countries; Fossil energy which will continue to provide for the majority of our needs; Finally, in spite of national efforts, CO 2 emissions which will probably be greater than the Kyoto objectives. Mtep 16,000 14,000 There are two conditions for this: firstly that we know how to respond to public opinion concerns. Then that we are capable of proposing new nuclear systems, even more effective in terms of safety or economy, but, above all, that will place in highest priority the sustainable development and non-proliferation criteria. But, making nuclear power acceptable is above all demonstrating it with proof. From this point of view, the exemplary operation of nuclear reactors for over 15 years, throughout the world, is an invaluable advantage. The availability rates are excellent, incidents, even minor, are decreasing and this enables public confidence to be gained. In the last few years waste management has been seen as the main problem of nuclear power for public opinion. It alone probably explains part of the defiance regarding nuclear power so well that it may have no future if we do not provide solutions for it. That said, contrary to the idea often spread, technical solutions do exist In France, as in other countries in fact, the management of less active waste and of that which has a shorter lifetime, is a reality already implemented in industrial disposal centres. It must be remembered that this represents more than 90% of the overall volume of nuclear waste. Source: AIE, World energy Outlook ,000 10,000 8,000 6,000 4,000 2, Fig. 1. A global production of energy with 87% of fossile origin! Renewables Hydrogen Nuclear Gas Oil Coal The question of high level and long lived waste, that which, with a few percent of the volumes, concentrates most of the radioactivity, remains. For this waste, the R&D employed in France, as outlined by law, has enabled numerous results to be obtained. This R&D will allow, by the legal deadline, in 2006, various technical solutions to be proposed to the French Parliament for the management of this waste. These studies also show that beyond 2020, even more than today, the environmental impacts will have to be studied with great urgency. Our first objective is to reduce the quantity of waste produced at source. Nuclear energy has many advantages for being a satisfactory energy response, in the long-term, from the resources and environmental point of view. We believe that it will have in the future, its place in an energy mix, even more than today. Nuclear energy of the future: what research for which objectives? 7

6 This involves evaluating and establishing the feasibility of processes enabling the waste to be separated, then the quantity and the noxiousness to be significantly reduced: this is partitioning* and transmutation*. The objective is to reduce the radiotoxicity of waste by a factor 100 by recycling plutonium and transmuting minor actinides. France has already chosen to recycle plutonium. Firstly because it is a recyclable energy material. Also, over time, it is the main element responsible for the radiotoxicity of waste. When exiting the reactor, spent fuel contains only 4% bona fide waste and 96% uranium and plutonium. In constant progress, the processing of spent fuel, then its recycling are solutions already implemented industrially in France. In this field of recycling, the research in progress targets the development of new fuel assemblies which will enable plutonium to be multiple-recycled, either in current water reactors or in EPR reactors, the deployment of which is envisaged in France.This would enable the stocks of plutonium with the current reactors to be stabilized, or even reduced. Once the plutonium recycled, a logical follow-up consists of partitioning then transmuting minor actinides (Np, Am and Cm) which are, after plutonium, the main contributors of waste radiotoxicity.the partitioning processes, developed in the wake of that which is currently being carried out for plutonium, will enable these minor actinides to be partitioned from fission products, thus considered as the sole final waste to be vitrified. As for the transmutation, its scientific feasibility is acquired, but its technical feasibility remains to be demonstrated and CEA is working on it, in an international, mainly European and American, collaboration. This then involves, for final waste, proposing technical solutions, which enable long-term waste management, either by storage, or by permanent disposal. The work carried out on conditioning* must enable processes to be proposed guaranteeing sustainable containment and the possibility of reusing the waste in complete safety, within a long-term storage* or deep geological formation disposal* perspective. Finally studies are being carried out on the long-term storage processes or on the disposal in deep geological formations integrating reversibility requirements. There again, we are starting to gather results: intact glass at 99.9% after 10,000 years, new storage concepts or even a new research laboratory for deep geological disposal. In France, these studies are carried out in compliance with the legal ethics and decisions which were made in 1991, in order to clarify by 2006 the Parliamentary and Governmental decisions. It is therefore up to the politicians to decide. But on a technical note, we will know how to deal with the small quantities of waste in question, and their low volumes originating from energy production, in safe conditions, in disposal or storage areas, for extremely long periods and assuring the traceability of any necessary information. Objective: To assure waste The advantages of this partitioning/transmutation strategy are very clear: it enables considerable reduction of waste radiotoxicity over the long-term. Thus, if the radiotoxicity of uranium used to produce fuel is taken as a reference, the same level of radiotoxicity is reached: After several hundreds of thousands of years if the fuels are not reprocessed; After 10,000 years if Pu is reprocessed/recycled according to the current solutions; After some hundreds of years if only fission products are left in the glasses and if the actinides are recycled. Directions selected Technical solutions Results 1. Conditioning Matrices and containers 1. Intact glass at 99,9 % after 10,000 years 2. Storage Durability 2. First storage concepts and/or Reversibility 3. Disposal Flexibility of the solutions 3. ANDRA research laboratory in construction Fig. 2. A major stake and realistic solutions. At present, various waste management strategies may be implemented complementarily. Reversible direct disposal, as will be the case in the USA with Yucca Mountain, storage in view of recycling at a later date, for example to give the system flexibility, or even immediate reprocessing and recycling, as is the case in France. According to CEA /SACLAY, Nuclear Energy Directorate. 1. For the meaning of all technical terms, refer to the developed glossary situated at the end of work. The terms in fat accompanied with an asterisk send back to the glossary (pp ). [Note of the Editor.] From now on, the recycling of all actinides seems to be a strong specification for reducing waste and proceeding toward sustainable development in nuclear energy. This criterion for 8 Introduction

7 reducing waste is furthermore widely repeated in the definitions for the designs of nuclear systems of the future. Today s challenge is imagining the nuclear systems of the future. With a first question: what future are we talking about? The objective is to develop systems that can be deployed from the industrial point of view by There are two reasons for this: firstly time is needed in order to propose truly innovative systems. If improvements in safety and competitiveness are expected, it is in fact technological ruptures that we are speaking of in terms of fuel, cycle and reactor core*. Then, it is the date on which the studies show an increased inflection of a requirement to use nuclear power with in particular the response to the electricity requirement but also the production of hydrogen*, the desalination* of seawater, etc. There already seems to be a convergence of opinions on the international level regarding the criteria the nuclear systems of the future must meet. These criteria, which privilege sustainable development, determine the order in which it will be necessary to try and decide on the research priorities. The CEA undertakes to work on three of them in particular: The sodium-cooled fast neutron concept, on which CEA already has a great deal of experience regarding reactors but which requires improvements regarding fuel cycles; The very high temperature gas-cooled and thermal neutron system for the production of hydrogen (VHTR); The gas-cooled and fast neutron system (FNR-G), which offers a promising alternative in relation to the sodium, regarding both reactor and cycle. On sodium reactors, the CEA has an important R&D programme in partnership with countries such as Japan and Russia. We are attempting to take advantage of the experience gained and the advantages of sodium, whilst improving the system on the difficult points. Various concepts of gas reactors were studied in the 70s-80s. Since, considerable progress has been accomplished in particular in the field of high temperature materials. Our ability to obtain high temperatures, and therefore high yields, place these reactors at the forefront. The R&D in progress concerns the materials, helium technologies, and the modelling supporting the developments. Economy Economy of natural resources Safety Reduction risk of proliferation Reduction of waste Fig. 3. Nuclear systems of the future: the 5 basic criteria. According to CEA /SACLAY, Nuclear Energy Directorate. One part, mostly dedicated to the VHTRs, concerns the materials for the high temperatures, the exchangers and the thermo-chemical cycles. The research centred on FNR-Gs will focus on the highly innovative fuels for these reactors. Nuclear energy will without a doubt play an important role in the future in order to meet international energy requirements. This, however, presupposes that the decision-makers know how to find and implement the correct responses to the question of waste, and to best take account of the sustainable development criteria. Research on this nuclear power of the future is developing within a largely international framework. For example, ten countries plus the European Union are participating in the American Generation IV initiative. This international work has already defined by consensus the most promising nuclear systems and drafted a common research and development plan for these systems. It will be necessary to innovate and redouble efforts in order to propose new concepts. This is carried out in a totally new framework which seeks to promote international cooperation, the sharing of tasks and results, and this within the group of countries who believe in the future of sustainable nuclear power Among the six concepts which have been selected after two years of preliminary work, the majority have a closed fuel cycle* and most have a fast neutron* core. This is the result of the sustainable development, waste reduction and optimisation and use of natural resources criteria. Nuclear energy of the future: what research for which objectives? 9

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9 The origins of current civilian nuclear power Fifty years ago, in December 1953, right in the middle of the cold war, the Atoms for Peace speech by the American President Eisenhower before the United Nations, prompted a deep mutation in the role of nuclear energy, until then limited to military usage. The president promoted its development for civil and peaceful use in order to serve the needs rather than the fears of humanity. The following year marked the start of the commercial generation of nuclear electricity in Russia. These initiatives have influenced energy policies because over these past fifty years, nuclear energy has developed widely throughout the world: reactors were in operation at the end of 2004, representing approximately 360 GWe installed in more than 30 countries.the proportion of nuclear power in the generation 250 of electricity is 16% (30% in countries of the 200 OECD), which also represents 7% of the primary energy. The first generation of reactors includes first prototypes constructed mainly in the United States, Russia, France and Great Britain. This first generation, developed in the 1950s-1960s, operated with natural uranium as enriched uranium was not yet commercially available. This is why during this period France developed the system called Natural Uranium Graphite Gas. It was then the Generation II of reactors which was deployed in the 1970s to the 1990s and which corresponded to most of the fleet currently in operation throughout the world. This generation arose from the need occurring in the 70s to make nuclear energy competitive and to reduce the energy dependence of certain countries at a time when considerable tensions were felt on the fossil energy market. Capacity in GW Ordinary water reactors, dominant species It is necessary to underline the industrial feedback, during these past decades, from all of these second generation reactors, which currently capitalize over ten thousand years of operation: it has in particular enabled the performances of the nuclear energy production to be demonstrated with a very competitive kilowatt-hour cost in relation to that of fossil energy. 0 BWR Boiling water Heavy water Graphite-Gas Water-Graphite Fast Total (HWR) GCR (Chernobyl) neutrons GW in GW in Fig. 4. Two main types of water reactor coexist: pressurized water reactors (PWR) and boiling water reactors (BWR). In the first, water from the primary circuit is under high pressure, which maintains it below boiling point although the temperature is significantly above 100 C; in the second, the pressure is lower, and the water boils on contact with the fuel. This period was that of the deployment of pressurized water reactors (PWR) and boiling water reactors (BWR), which together currently constitute over 85% of the global electronuclear fleet. Overall, this industrial maturity, this satisfactory competitiveness and this favourable feedback have contributed considerably to renewing the electricians confidence in nuclear energy. The ready availability of their power plants and the possibility, for some of them, to see their lifetime extended up to 50, even 60 years, strengthens this trend. Nuclear energy of the future: what research for which objectives? 11

10 Plants installed 1 plant in construction (execution orderrrr given) Units permanently shutdown (11 units) Penly Gravelines Chooz Flamanvile Paluel Cattenom Mont d Arrée Reactor system Natural Uranium Graphite-Gas Gaz-Heavy water Breeder reactor PWR* open circuit cooled PWR closed/looped circuit cooler *Pressurised ordinary Water Reactors Chinon Saint-Laurentdes-eaux Le Blayais Civaux Golfech Dampierre Nogent-sur-Seine Belleville Saint-Alban Saint-Maurice Cruas Tricastin Marcoule Fessenheim Bugey Creys-Malville Phénix The neutrons produced by a fission reaction may induce new fissions of other fissile nuclei present within their vicinity, and thus contribute to maintaining the chain reaction. In a PWR, the water is both a coolant and neutron retarder. The water circulates across a forest of fuel assemblies, long bundles of thin zirconium alloy metal tubes, where the uranium oxide or plutonium ceramic pellets are stacked km Fig. 5. There are 59 reactors in France, producing a capacity of 63 GWe. France has replaced all of its first generation graphite gas reactors with PWRs. Given the lifetime of the reactors and the time necessary for developing new systems, water reactors will certainly remain preponderant in the global nuclear fleet up to 2030, and probably during all of the first half of the 21 st century. Source : EDF This water which circulates in a very thick steel closed circuit, yields its calories by making the water of a secondary circuit boil into a steam generator. The steam thus produced will activate the turbo-alternator. After being distributed in the turbines, the steam is condensed by way of a new water circuit, itself in thermal contact with a cold source, atmosphere, river or sea. The operation of a pressurized water nuclear reactor A pressurized water reactor is none other than a developed device designated to heat water, with inside the boiler a pressure of 150 bars and a temperature of 300 C. The principle of such a reactor is to permanently maintain fission reactions of the uranium or plutonium nuclei within an environment, called reactor core*. Each fission, induced by the neutrons present in the core, releases energy in the order of 200 MeV*, and produces two or three additional neutrons, one of which serves to maintain the chain reaction*, the others being absorbed in (the water or) the structures or lost outside of the core. Free neutron Fissile atom γ radiation γ radiation Fission product Freed neutron Fission product A pressurized water reactor is from the group of reactors, called thermal neutron, that is the high energy neutrons produced by the fission are slowed down by successive shocks in an environment that is called a moderator*, in order to obtain thermal equilibrium with this environment. They therefore have a much higher probability of inducing new fissions. Fig. 6. Nuclear fission: under the impact of a neutron, a heavy nucleus such as uranium 235 may fission, and provide two lighter nuclei (fission products) and a few neutrons. The reaction releases energy 200 million times higher than that typically called into play in a chemical reaction between atoms or molecules. 12 The origins of current civilian nuclear power

11 Fig. 7. The neutrons produced by a fission reaction may induce new fissions of other fissile nuclei present within their vicinity, and thus contribute to maintaining the chain reaction. Primary circuit Vapour generator vapour water Secondary circuit Pressurizer Turbine primary pump Cluster command mechanisms Generator Reactor core Vessel Water supply pump Condenser Reheater Collant water Fig. 8. Diagram of a pressurized water reactor (PWR). Nuclear energy of the future: what research for which objectives? 13

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13 The fuel and the fuel cycle Nuclear fuel is designed to supply the power expected from the reactor, by best using the fissile material. The design of the fuel element must in addition allow a certain flexibility in the reactor operation, in order to enable it to adapt to the variations in power imposed by the system. This must be carried out without releasing radionuclides* from the nuclear reactions into the reactor s primary circuit. These constraints combine in fact to give the nuclear fuel leaktightness, robustness and reliability qualities. The fuel assembly of an ordinary water reactor is always made up of fuel rods* containing the nuclear materials, arranged in a square lattice array in a structure assuring in particular the mechanical maintenance of the rods. The fuel rod is made up of pellets of uranium oxide or mixed uranium and plutonium oxide (diameter and height in the order of 1 cm) stacked in metal tubes (cladding* in zirconium alloy) sealed at the ends (leaktight). Plug Orifice for pressurisation Upper plug Spring UO 2 pellet Zircaloy cladding Lower plug Fig. 10. Fuel rod for a PWR reactor. Control cluster Upper end fitting Guide-tube Mixing grid Fuel rod Lower end fitting Fig. 9. UO 2 fuel pellets. Fig x 17 Fuel Assembly and control cluster. The robustness and reliability of the fuel must enable a long stay in the reactor (currently 4 years, with an objective of 6 years towards 2010 for French reactors). The integrity of the cladding is very important because this is what constitutes the first barrier* between the radioactive products and the environment. The fuel rod s cladding must Nuclear energy of the future: what research for which objectives? remain leaktight in incidental or accidental situations, even at the end of the fuel rod s life. That said: Some fission products are gaseous: their production progressively increases the pressure inside the cladding; The chemical composition of the pellets is modified by the appearance of fission products and actinides; 15

14 The fuel ceramic swells under irradiation and imposes a stress on the cladding which contains it (pellet-cladding interaction). Only the odd Pu and U 235 isotopes are fissile to thermal neutrons. The irradiation of the U 238 form of Pu. In a water reactor, a certain number of neutrons are absorbed by the water: it would therefore be impossible to maintain the chain reaction if natural uranium, which only contains 0.7% fissile 235 isotope, was used for fuel. It is therefore necessary to enrich the uranium, up to a content of approximately 4% of U 235 (see inset). The fuel is taken out of the reactor when it no longer contains enough fissile nuclei in order to maintain the chain reaction (typically at the end of four years in a water reactor). After its stay in the reactor, the fuel no longer contains enough fissile material to maintain the chain reaction, but it is not necessarily exhausted. As shown in the diagram above, it still contains a large quantity of fissile and fertile* material, which is important to recover. It also contains fission products and minor actinides which make it extremely radioactive and difficult to handle. The finality of the reprocessing is double: Recovering the recyclable energy materials; Separating these materials from the true waste, and conditioning* the latter in an inert and safe form (vitrification*). 4 U 235 U Uranium 238 In France, these operations are carried out in the Cogema plant IN La Hague. Uranium REP UOX Plutonium Pu 3 2 Minor actinides 5 0,1 PF 1 1 U Fission products Fig. 12. Reactions within the standard fuel assemblies in the PWRs (45,000 MWd/t). The combined play of fissions and neutron captures in the fuel of the water reactor may be summarized as follows (see figure below): we start with 100 uranium atoms, four of which are isotope 235 (fissile) and 96 isotope 238. Of the four, only one will survive, and three will undergo fission. Of the initial 96 U 238, three will be transformed into Pu and 93 will survive. Of the three Pu formed, two will undergo the fission and only one will survive. In total, there will have been 3+2 = 5 fissions: only 5% of the heavy metal will therefore be consumed in a water reactor. In a fast neutron reactor*, the schema would be very different with a greater consumption of the fertile isotope U 238. Fig. 14. The COGEMA reprocessing plant in La Hague, in which the spent fuel reprocessing and waste conditioning operations are carried out. Most of the radiotoxicity of the spent fuel comes from plutonium.this is an additional reason for recycling it and not leaving it in the waste. U 470 kg (94 %) Recyclable materials Pu 5 kg (1 %) A.M. 0,7 kg (0,15 %) Final residue P.F. 25 kg (5 %) Fig. 13. Composition of a 500 kg enriched uranium assembly after its passage in a reactor. 16 The fuel and the fuel cycle

15 Reprocessing plant U Pu Recyclable materials Uranium PF AM Structure waste Technological waste Plutonium Final waste Vitrified residue Compacted waste Cemented waste Fig. 15. The streams through the reprocessing plant. The nuclear fuel cycle It is not the uranium ore directly which constitutes the nuclear fuel. So that the heavy nuclei can be used in a reactor, they must follow a fuel cycle* which combines numerous industrial stages: Extraction of the uranium ore; Concentration of the ore; Conversion of the uranium concentrates into gaseous uranium hexafluoride (UF 6 ); Isotopic enrichment* of the uranium in the UF 6 form, in order to increase the proportion of U 235 fissile nuclei, too low in natural uranium; Manufacturing of the fuel (conversion of the fluoride into enriched uranium oxide UO 2, making into pellets, pellet sintering*, rodding, assembly of rods into bundles). The fuel produces energy for approximately four years in the reactor. The last stages are therefore: Interim storage, under water, of the spent fuel; Management of the spent fuel. This stage differs according to what is considered as a closed or open cycle. The open cycle, which is not really a cycle, ends by the final disposition of the spent fuel, therefore considered in block as waste*.the open cycle is currently practized in the United States, Sweden, etc. Natural uranium Depleted uranium Storage Conversion Concentration Extraction minerai Enrichment The closed fuel cycle is the one practized in France, Germany, Switzerland and Japan. The following substages are found: Chemical reprocessing of spent fuel in order to recover the fissile and fertile materials that it still contains, in view of recycling them; Recycling of the plutonium in the form of MOX* fuel (acronym for Mixed OXide fuel); Conditioning of the waste, and, in particular, vitrification of highly radioactive waste resulting from the fission; Final disposition of the conditioned waste. Each cycle facility, enrichment, manufacturing, or reprocessing plant is dimensioned is in order to supply several dozens of large reactors. Enriched uranium Recyclable uranium Permanent disposal Manufacture of fuel UO 2 fuel MOX fuel New MOX fuel Plutonium Reprocessing Final residues Spent MOX fuel New UO 2 fuel Reactor Spent UO 2 fuel Fig. 16. The nuclear fuel cycle. Nuclear energy of the future: what research for which objectives? 17

16 The enrichment of uranium An important stage of the cycle is uranium enrichment. Isotopic partitioning is a very difficult undertaking because the isotopes to be separated have the same chemical properties and almost the same physical properties. Two main enrichment techniques are industrially implemented throughout the world: Gaseous diffusion, which consists of passing uranium, in the gaseous UF6 form, into a porous medium by exploiting the fact that the light isotope diffuses a little more quickly than the heavy isotope. The elementary process enriches very little, which obliges the operation to be repeated a great number of times in succession in order to obtain the suitable level of enrichment. Why recycle plutonium? MOX fuel Today, plutonium is recycled in water reactors, PWRs and BWRs which constitute the main part of the global electronuclear fleet. This enables saving enriched uranium, for which plutonium is substituted in part, and preventing the plutonium ending up in final waste or only accumulating on shelf after having been partitioned during the reprocessing of spent fuel. This recycling is carried out in MOX fuel. The reprocessing/ recycling combination also enables the quantities of spent fuel stored in pools to be significantly reduced. A MOX fuel, made up of a solid solution of plutonium and uranium oxides, is outwardly in every way identical to the enriched uranium fuel that it replaces. The pellets which fill the cladding have identical dimensions: only their composition and their manufacturing process change. In the core of a water reactor, due particularly to the presence of non fissile plutonium isotopes, it is necessary to place approximately twice as much plutonium in order to obtain the energy equivalence of an assembly enriched with U 235: in order to replace the uranium enriched to 4%, a mixture containing approximately 8% of plutonium and 92% of depleted uranium will be necessary. At the end of its life, the MOX fuel will contain no more than approximately 4% of plutonium. There is thus a net consumption of plutonium: the use of MOX enables the increase in the plutonium inventory to be limited in the fleet of reactors. Fig. 17. The Georges Besse enrichment plant, in Pierrelatte. Expensive in energy, gaseous diffusion is currently being replaced by ultracentrifugation, which consists of making the UF 6 gas circulate in a centrifuge rotating at very high speed. The heaviest molecules concentrate on the periphery, which enables the two isotopes to be partitioned. As each centrifuge has a low material flow rate, this technology therefore requires many centrifuges to be working at the same time. The recycling of spent fuel in the form of MOX began experimentally in Belgium at the beginning of the 60s. It was then industrialized in this country, in Germany and in Switzerland, then in France from Today, Japan is preparing, in turn, to MOX BWRs and PWRs, and the United States is seriously thinking about it. In France, EDF decided to recycle its plutonium progressively in some of the reactors of its fleet. The 20 MOX reactors recycle all of the plutonium effectively extracted by reprocessing EDF fuel at the UP2-800 plant in La Hague. The plutonium inventory of an MOX PWR is balanced: as much plutonium is produced in the enriched uranium fuel bundles as is consumed in the MOX bundles. The economic profitability of MOX depends a great deal on the authorized irradiation rate, that is the overall quantity of energy that a given fuel may supply, hence the research, currently carried out, which aims to increase this rate. No fundamental obstacle opposes a long irradiation time for the MOX, because the behaviour of MOX assemblies in reactors is very similar to that of uranium fuels. Fig. 18. A succession of centrifuges for the enrichment of uranium. 18 The fuel and the fuel cycle

17 Radioactive waste and its current management Origin of radioactive waste When a neutron causes the fission of a heavy nucleus, it splits into two unequal pieces. These fission fragments are rarely stable nuclei. Apart from fission products, the neutrons induce the formation of actinides and activation products, originating from their neutron capture by non-fissile nuclei, radioactive species that are partly found in waste. The radioactive decay of these various species may be, according to the case, fast, slow or very slow, from fractions of microseconds up to billions of years. These various species, partitioned during reprocessing, constitute the main source of high level and long lived waste. However, throughout the fuel cycle and during the operation of the reactor, inert materials are contaminated by radionuclides resulting from nuclear reactions in reactors. These are carefully isolated and conditioned and constitute another category of waste, called low or intermediate level, much less radioactive but more abundant. In this category, waste contaminated by radionuclides but which has another origin than the electro-nuclear industry and which are caused by conventional industry, research or medicine are also found. Category B: low and intermediate level (A) long lived waste (several thousands of years and more). Example: the spent fuel rod cladding segments, after dissolution of the fuel itself during reprocessing. Category C: long lived waste and high level waste, emitters of α, β and γ radiation, release heat for several hundreds of years and remaining radioactive much longer. It concerns either unprocessed spent fuel (four countries having renounced reprocessing ), or glass containers from the reprocessing and which incorporate fission products and minor actinides. In France, rather than speak of A, B or C, ANDRA* and the Nuclear Safety authority thus classified waste, according to the system implemented for their long-term management: The various categories of radioactive waste Low level LL Centre de stockage Dedicated disposal For its daily management, radioactive waste is de l Aube under review classified according to two criteria: The level of activity*, that is the intensity of the radiation that it emits, which conditions the importance of protections to be established, in order to protect ourselves from radioactivity; Intermediate level IL High level HL A waste C waste B waste C waste The radioactive half-life* of the products contained, which What quantities? enable the duration of its potential harmfulness to be defined. Thus in general three categories of radioactive waste are distinguished. Category A: low and intermediate level short lived waste (radioactive half-life less than 30 years). Its radioactivity (β and γ) will be reduced to a level comparable with natural radioactivity between now and 300 years. It may come from power stations and fuel cycle plants, but also from hospitals, laboratories, industry, etc. Short lived Long lived Very low level VLL Disposal at Morvilliers Secured for (Aube) since 2003 mine tailings In France, where three quarters of electricity is nevertheless produced by nuclear energy, the quantities concerned represent less than 1 kg of radioactive waste per inhabitant and per year, that is 0.04% of the industrial waste (2,500 kg/inhabitant/year). This quantity is distributed as follows: 900 grams of A waste, which however only contains 5% of the overall radioactivity; 90 grams of B waste; 10 grams of C type. Nuclear energy of the future: what research for which objectives? 19

18 The annual production of B and C waste, from the reprocessing in La Hague of spent fuel from French reactors, is in the order of 700 m 3 per year, of which less than 200 m 3 /year for glass. B and C waste is currently stored in La Hague (but the waste from the reprocessing of foreign fuel is sent back to its owners). The ultimate future of long lived waste ( B and C ) m 3 /t Fig. 19. The centre de stockage de l Aube, for category A waste (short lived). The overall production of A waste (packages included) is approximately 15,000 m 3 /year, a volume which regularly decreases thanks to the efforts of the producers. After conditioning, the packages are sent to the Centre de Stockage de l Aube (CSA). ANDRA stacks these packages in reinforced concrete cells which, after filling and packing out the spaces with gravel or mortar, are sealed with a concrete slab and coated with a waterproofing polymer. Lastly, a leaktight cover will be placed and the site covered with a few metres of earth. The overall capacity of the CSA is 1 million m 3, which, at the current rate of production of this waste, assures it an operation at least until Design values Reduction technological waste Bituminisation stop Compacting commissioned 1991 values 1995 values 2003 values Source: ANDRA. Fuel disposal in status (estimation) A seemingly necessary solution From the beginning of nuclear power up to the 80s, most specialists shared a common vision of the final management, the disposal of highly radioactive waste in deep geological strata: in order to permanently isolate it from the human environment (today known as the biosphere), it would be buried in a leaktight way, deep enough, in a fairly stable geological stratum, isolated by judiciously arranged engineered barriers. Under these conditions, the time necessary for the radionuclides contained in the waste to migrate up to the surface, after corrosion of the packages by ground water, would largely exceed the time necessary for the radioactivity to decay and return to a natural radioactivity level. Conditioned spent fuel Bitumines Technological waste Structural waste Glass Almost all of the countries equipped with reactors have studied variants of this same solution, according to the geological nature of their subsoil and the respective qualities of the stratum envisaged: salt, clay, granite, basalt, etc. Throughout the entire world, approximately fifteen underground laboratories have been installed to study on site the characteristics of the stratum which would host the waste and the behaviour of the geological barrier. The main topics of study concerned and still concern the mechanical resistance of rocks, the network of faults, the physical chemistry and the flow rate of the ground water, the mechanisms and the degradation kinetics of the packages, etc. However, in spite of this research, no long lived and high level waste disposal has yet been implemented in the Western world 2. In France, a law enacted on the 30 December 1991 prescribed the continuation of research on the long-term management of long lived and high level radioactive waste. Fig. 20. The volume of conditioned waste regularly decreases thanks to the efforts of producers. 2. A deep disposal site of long lived intermediate military waste was opened and commissioned in the USA, in 1999 on the WIPP site (New Mexico). 20 Radioactive waste and its current management

19 The steering of the research relating to lines 1 and 3 is entrusted to CEA, that of the research for line 2 is a matter for ANDRA. Fig. 21. Storage of vitrified waste on the COGEMA site at La Hague. This research has mobilized the entire French nuclear scientific community and takes advantage of internationally accumulated knowledge. It is declined according to three main points: Line 1 concerns the methods of enhanced partitioning of waste from very long lived radionuclides and the possibilities of their transmutation by nuclear reactions into shorter lived species, or even, ideally, stable nuclides*. The reprocessing of spent fuel is a mandatory prerequisite for any partitioning/ transmutation; Line 2 concerns geological disposal and involves the construction of underground laboratories to study on site the formations presumed as favourable; Line 3 concentrates on the conditioning of waste in view of enabling, if necessary, its storage in complete safety over long periods. Line 1 programmes have permitted the definition and experimentation on the laboratory scale of enhanced partitioning processes, as well as a certain number of experimental demonstrations of the physical feasibility of the transmutation of certain long lived radionuclides. Partitioning/transmutation offers the prospect of considerably reducing the quantities to be disposed of, and therefore the cost of disposal, but final waste will always remain. This leads to the belief that geological disposal will be necessary. Research on disposal results in the digging of an underground laboratory in a clay formation in the Parisian Region, on the Bure site, at the edge of Meuse and Haute-Marne, under Andra s responsibility. CEA is also associated with this research, in the capacity of main contractor or service provider for some experiments. Studies on the long-term conditioning of waste are being continued, with a large knowledge base between disposal and storage. A Commission Nationale d Évaluation (CNE) National Evaluation Commission, consisting of experts appointed by the government, follows the progress of this research and reports annually to the Parliament and the Government. Line 1 Enhanced partitioning Line 3 Conditioning and storage In 2006, at the end of these 15 years of research and according to their results, the national representation will once again take hold of the subject and will make the necessary decisions. Line 2 Reversible disposal Fig. 22. The three main lines of research on waste management. Nuclear energy of the future: what research for which objectives? 21

20

21 The decommissioning and dismantling of nuclear installations The issues Nuclear installations, whatever their nature laboratories, control or production plants, experimental or electricity producing reactors, radioactive waste reprocessing installations, etc. have a limited operating time. When their nuclear installations become old, many countries are led to shut down operation, decommission them and dismantle them. The end of life of a nuclear installation may be caused by the completion of experimental programmes planned in the installation, the obsolescence of materials and processes, economic (optimization of means, cost of maintenance) or safety and security (change in the regulations) considerations. Decommissioning and dismantling* (D-D) aims to enable the partial or total liberation of a nuclear site. Three stages may be distinguished for the decommissioning of a nuclear installation: the permanent closure, the decontamination-dismantling, then the demolition and liberation of the site. In the case of a reactor, the spent fuel is removed from the core and stored or reprocessed. The circuits are drained, the operating systems switched off and the openings to the exterior locked and sealed. The containment atmosphere is checked and the access to this containment is restricted; monitoring systems are installed. In general, permanent closure intervenes very shortly after the permanent shutdown of the reactor. Then the decontamination of the surfaces of the buildings and the material takes place. Decontamination techniques serve to reduce the installation s radioactivity, to clean up the metals and the concrete in the aim of facilitating the access to the work areas and the handling of the elements and material to be dismantled, to enable the cutting work and to meet the standards regulating the evacuation of waste. All of the operating equipment is dismantled and, after checking its residual radioactivity, recycled or temporarily stored. Only the reactor s structures, in particular, the vessel and its protective shielding, are left on site. To finish, in a third stage, all of the remaining materials and the installation itself will be dismantled then the site decommissioned and liberated for other uses. In some cases, a very long timeframe may pass, which may reach several decades after the shutdown of the installation, up to this final stage.this Nuclear energy of the future: what research for which objectives? long timeframe enables radioactive decay and therefore easier protection of the workers who proceed with the deconstruction operations. It also facilitates the storage then the final disposal of the radioactive waste. Decommissioning-dismantling: one of CEA s important issues A historical player in nuclear research in France, the CEA must manage the heritage of the past. It first involves the recovery work and conditioning of old waste. CEA also has numerous installations of various types to dismantle in its own centres. Cleanup and dismantling actions from now on constitute one of the important requirements of CEA s policy. Since the beginning of 2002, the part of the actions that can be attributed to catching up with the past is covered by a dedicated fund taken from CEA s participation in the Areva group. In particular, this concerns the management of old waste (the production of which is prior to 1992), spent fuel and useless radioactive sources, the dismantling of installations placed in permanent shutdown, the cleaning up of the environment, the construction of service installations and the manufacturing of transportation packaging relating to these actions. The use of the dedicated fund is controlled by a Monitoring Committee, whose members represent in particular CEA s sponsoring departments. The R&D requirements in the nuclear field and the orientations taken in order to meet them lead CEA to grouping the majority of the experimental nuclear installations in operation in Cadarache and Marcoule in the not so distant future (in the order of 10 years). The number of installations to be treated (approximately thirty installations from 2001 to 2010) means that CEA s decommissioning-dismantling programme is very large in volume. In 2012, CEA will have completed the dismantling and radioactive cleanup of the Fontenay-aux-Roses site and in 2015 that of the Grenoble site installations. Only the LECI* hot laboratory, for the programmes concerning materials and structures, and the Orphée reactor, for basic research programmes, will remain in Saclay. 23

22 Cadarache, which will group between now and ten years a large part of the experimental nuclear installations, must be able to manage all of the waste produced on the site locally and offer its capabilities to other Centres, for waste which is not their own. It is therefore at Cadarache that investment efforts regarding service installations are concentrated. The decommissioning-dismantling: an important emerging market Decommissioning practices are realizing their full potential and may be considered from now on as a controlled phase of a nuclear installation s life cycle. A few figures illustrate the scope of the commercial D-D issues: more than 500 nuclear power plants have already been constructed and operated throughout the world, among which 108 were decommissioned in January In addition to the power plants, there are also plants connected to the manufacturing of fuel and reprocessing of spent fuel, a part of which has already been or will soon be decommissioned. As a general indication of the overall level of the D-D costs, the United States regulating body demands that operators have at least 164 million dollars (2000 value) in order to decommission and dismantle a conventional pressurised water reactor. The average age of nuclear power plants in OECD countries is approximately fifteen years, in relation to an average effective lifetime of at least 30 years. The decommissioning rate should peak around France is characterized by the large number (six) of graphitegas reactors which have been shut down, and by the number of R&D and demonstration power plants currently shut down. Fig. 24. Institut national des radioéléments Fleurus/Belgium: decontamination by gel of the C2 cell (XEMO I chain). Final status. Existing and future technologies Dismantling techniques already exist and installation design and decommissioning projects benefit from large amounts of feedback. In general, decontamination techniques call upon chemical, mechanical or thermal processes, or a combination of these. In order to decontaminate concrete or metal surfaces, the very high-speed projection of dry ice granules and the use of chemical gels or decontaminating foams are used for example. Dismantling calls upon cutting techniques for metal or concrete structures. Mechanical (such as sawing or high pressure water jet) or thermal (plasma torch) processes are used for example. Radiological measurement techniques are used to address the inventory of radioactive stocks in the installation, sort the materials and waste according to their category, and to take the necessary provisions to protect the workers. Photo STMI. Dismantling uses various techniques: removable shielding, temporary airlocks and cells, mobile filtering and ventilation systems, special clothing, ventilated protective suits and masks. Dismantling also uses lifting and handling equipment, and makes extensive use of remote control techniques: remote handlers, semi-automatic tools enabling the employees to work at a certain distance from the sources of radiation. Fig. 23. Projection of pressurized foam. 24 The decommissioning and dismantling of nuclear installations

23 Decommissioning-dismantling feedback Numerous nuclear installations have already been successfully decommissioned and dismantled. Here is the list of installations dismantled or in the process of dismantling in France: Fig. 25. Cutting an auxiliary boiler in a ventilated flame resistant suit. Dismantling of the Brennilis EL4 power plant. The EDF media library/antoine GONIN. Power reactors - The Monts d Arrée power plant (EL4). - Natural uranium graphite gas (NUGG) system reactors. - The Chooz A D reactor (Ardennes nuclear power plant). - The Superphénix reactor. Research reactors - The Rapsodie reactor. - The Harmonie reactor. - The Mélusine and Siloé reactor. - The Strasbourg University reactor. Dismantling waste The dismantling of nuclear installations produces a large quantity of waste, mainly of low level. The European Commission estimates that the decommissioning of an average nuclear power plant produces up to 10,000 m 3 of radioactive waste. Concrete and other construction materials only containing a very low radioactivity represent, in volume, the main part of this waste. The effective management and evacuation of radioactive waste is an essential condition of the success of the D-D of nuclear installations and represents the main part of the overall cost (in the order of 60%, or installations together, according to a German estimate). The large quantity of dismantling waste containing only very small concentrations of radionuclides requires special care in the reduction to the minimum of the constraints linked to their evacuation as radioactive waste. This leads to waste zoning of the installation being carefully carried out, by accurately defining the border between conventional waste areas and radioactive waste areas. Dismantling waste currently has, in France, a specific disposal centre (VLL centre, Very Low Level) in Morvilliers. CEA laboratories and workshops - The AT1 pilot reprocessing workshop. - The caesium 137 and strontium 90 source manufacturing workshop (ELAN IIB). - The enriched uranium reprocessing workshops (ATUE). - The fuel assembly cutting laboratory (LDAC). - The plutonium chemistry laboratory (LCPu). - The plutonium based fuel laboratory. - The Saturne accelerator. - The Saclay linear accelerator (ALS). The other installations - The FBFC plant at Pierrelatte. - The irradiator of the Société normande de conserve et stérilisation (SNCS). The assessment of the operations carried out show that until now, only small research reactors have been the subject of a total dismantling with complete deconstruction of the buildings. Medium-sized reactors (G1, G2, G3, EL3, Rapsodie) have only been subjected to partial dismantling, due to the absence of associated waste (graphite, sodium) disposal systems. Several laboratories, workshops or pilot plants have been completely dismantled. Finally, an ore reprocessing installation, which produced almost 10,000 tonnes of uranium in metal and oxide form, as well as thorium*, was completely dismantled. The analysis of these operations leads to the observation that the dismantling of reactors and fuel manufacturing installations (hot cells and plutonium laboratories) is considerably shorter than that of installations involving chemistry (ore processing, reprocessing) and contaminated by fission products. It has Nuclear energy of the future: what research for which objectives? 25

24 been noted on the operations carried out that the volumes of waste generated, of a few hundreds to several thousands of m 3, can be properly managed. Dismantlings in progress at CEA are good examples. The experience which will thus be acquired on small or mediumsized installations will certainly be very useful for the dismantling of large nuclear power plants or certain plants of the front or back-end of the civilian nuclear cycle.

25 Nuclear safety and security The design, construction and operation of nuclear installations must take account of the safety requirements, and their impact on Mankind and the Environment must be controlled. It is an essential link for the public s acceptance of nuclear energy. The risks associated with nuclear power are perceived by the public as important The acceptance of the nuclear risk poses a problem for society. In the medical field, the risk is accepted because it is balanced with estimated benefits. This is rarely the case when one becomes interested in energy production. That said, the French Académie nationale de médecine 3 observed that the most serious health risk is a lack of energy (link between health status and energy expenditure in developing countries, health consequences from supply shortages, etc.) and recommends maintaining the nuclear system in so far as it proves to have the lowest impact per kwh produced in relation to the systems using fossil fuels, biomasses or the incineration* of waste, or even when it is compared to wind and photovoltaic energies. The comparison of factual data between nuclear risks, other industrial risks, risks arising from other human activities (transport, tobacco, etc.), natural risks, etc. is instructive. But concerns regarding accidents, long lived waste and the impact on future generations cannot simply be dissipated.the perception of risks is eminently subjective; those that result from choice (e.g. rock-climbing), and those which result from equipment imposed by the community (nuclear power plants) are not perceived in the same way. The acceptance of nuclear power by the society passes in any case by a permanent communication and transparency effort (in particular regarding accidents and incidents), and by the independence of a strong supervisory authority for operators. Nuclear power and environment Radioactivity is found throughout the environment. But most of this radioactivity is of natural origin. It comes from cosmic rays, radon* from minerals from the earth and exhaled into the air that we breath, terrestrial radiation coming from the isotopes from the uranium and thorium chains present in the ground, and carbon 14 and potassium 40 present in our organism and in our food. However, we also find in certain compartments of the environment, artificial radioactive isotopes, originating from the atmospheric nuclear tests carried out during the Cold War, fallout from Chernobyl, and finally, for a very small part, from industrial nuclear activities. In normal operation, the environmental impact of nuclear installations is low: power plant emissions (tritium) are barely detectable (and yet, we know how to detect radioactivity at very low levels, but natural radioactivity easily masks the human induced contribution); emissions from the La Hague reprocessing plant are much higher and much easier to detect 4 (iodine 129 and tritium are discharged into the sea, krypton and tritium into the atmosphere). They also include a chemical discharge element (nitrates, marginal compared to the agricultural contribution). However, the effects of dilutions and dispersion in marine or atmospheric environments render the radioactive contribution from the plant insignificant compared to the natural contribution a few kilometres away from the installation. All radionuclides do not behave in the same way. Their behaviour depends on their chemical properties. In most cases, a dispersion and dilution of the contaminants is observed. In others, conversely, a concentration in certain compartments of the biosphere is observed. Reconcentration phenomena may be of biological origin (case of caesium in mushrooms) or have any physical or chemical cause (contamination* marks observed in Mercantour are caused by runoff phenomena). 3. Report concluding a colloquium held on June 25 th, 2003 on the relationship between health and the energy choices. Nuclear energy of the future: what research for which objectives? 4. The La Hague reprocessing plant discharged in 1997 approximately 12,000 terabecquerels (Tera = x 10 12, ie multiplied by a million of millions) in the form of liquid waste (11,900 TBq of tritium and 1.8 TBq of iodine 129) and 300,000 terabecquerels in gaseous form (mainly krypton 85). 27

26 Rain Atmosphere Ground deposit Infiltrations Gaseous emissions Ground waters Atmospheric diffusion transfers Rain Milk Foliage Plantations Stream Toward drinking water Watering Inhalation External exposure Ingestion Drinking water Liquid effluents Sediment Hydrological/ hydrogeological diffusion transfers Fig. 26. The analysis of the transfer of radioactive contaminants into the biosphere is the subject of radioecology*. Different compartments of the biosphere are considered: soils, lakes, rivers, atmosphere, plants, animals and humans. Ingestion Fish Algae Atmospheric fallouts (aerosols) Bathing Water sports Shellfish Marine diffusion transfers (decree of March 2001) for public exposure. As a guide, the average natural irradiation is 2.5 msv/year, but it is necessary to specify that the abovementioned dose limits concern doses in addition to human activity. In comparison to natural doses, the dosimetric impact of nuclear installations is low. A nuclear power plant discharges into the environment ten times less radioactivity than a coal or fuel power plant of the same power: the collective dose is between 1.6 and 2.6 man-sievert per gigawatt-year for a nuclear power plant, compared to 20 for a coal power plant. The impact of cycle plants (reprocessing, mines) is much greater: according to the last report from the Nord-Cotentin Commission, the dose induced by discharges from the La Hague plant on the most exposed population is 0.06 millisievert per year, that is 20 times less than the dose caused by natural radioactivity. Nuclear power and health risks We are all exposed to natural radioactivity, and the effects of radioactivity on the organism are no different depending on whether the radioactivity is of natural or artificial origin. If the effect of high doses resulting from serious accidental situations is well-known, the problem of low doses of radiation remains a subject of biological and medical research (relationship between the risk and the dose, threshold effect), with an epidemiological section. The same goes for the hereditary effects of radiation. The control of the exposure to radiation is the subject of radiation protection. Current French regulation imposes a dose* limit of 20 millisievert (msv)* over 12 consecutive months (decree of March 2003) for worker exposure, and 1 msv/year Scale of risk relating to effective annual dose Major health risk Inhalation (1,2) Natural origin External exposure Internal exposure Medical origin (diagnostic needs) Aerial nuclear weapon tests Chernobyl Nuclear industry Nuclear tests (0,08) Chernobyl (0,002) Medical origin (0,4) Nuclear industry (0,0002) Ingestion (0,3) Artificial origin Cosmic rays (0,4) Terrestrial gamma rays (0,5) Cosmic rays Terrestrial gamma rays Inhalation (mainly radon) Ingestion Fig. 27. Components of the annual radioactive dose (in msv). Source: UNSCEAR 2000 Reminder: 20 msv workers limit 1 msv public limit Range of natural exposure the most often encountered Normal situation 10 msv/year AND* 1 msv/year msv/year msv/year + 20 msv/year + 10 msv/year 2,5 msv/year -1 msv/year -µsv/year *AND = average natural dose 400 AND Significant health risk 40 AND Low health risk 4 AND Insignificant health risk zero or practically no health risk Fig. 28. Scale corresponding to the levels of exposure and health effects. Source: Clefs CEA, n 48, summer 2003, Chemical and radiological toxicology. 28 Nuclear safety and security

27 Safety and demonstration of safety In the nuclear industry as in any human activity, zero risk does not exist. The objectives of the safety procedure is therefore not to entirely eliminate the risks associated with nuclear activities. In a less ambitious but more realistic way, it is to pre-empt the risks of accident, and to mitigate their consequences in the hypothesis where the accident would nevertheless occur. The notion of safety is taken into account very early on, as of the installations design phase.the specificity of the nuclear industry arises from the use of radioactive materials, which are likely to be dispersed into the environment or even to affect the human being, and which are at the origin of ionising radiations with multiple effects (irradiation, thermal energy, radiolysis, etc.). French regulation mainly requires deterministic calculations (incidents or accidents are postulated). With the safety objectives defined, possible failures are imagined, which may be of external or internal origin, (earthquake, fire, power cut, pump shutdown, etc.), the behaviour of the installation is simulated, and it is made sure that the consequences are acceptable. All of the difficulty resides in the exhaustiveness of the list of scenarios envisaged. A set of principles, concepts and methods has been developed, both at the design stage, and at the construction or operation stage. Thus defence indepth* consists of interposing several lines of defence (follow-up of actions, equipment or procedures, grouped in levels each one of which has the purpose of pre-empting degradations likely to lead to the next level and to limiting the consequences of the failure of the preceding level) in relation to aggressions that may affect the safety functions. This is generally assured by the redundancy and the diversity of barriers (successive and leaktight multiple-barrier system). Several means of stopping the chain reaction, redundant and diversified residual power evacuation systems, several barriers between the radioactive products and the environment, thus exist. It is endeavoured to make these various means as independent as possible from one another, and to plan for each of them permanent or periodical monitoring used to guarantee their availability. Risk analysis is the subject of a conventional procedure: Technical analysis of the installation s safety and reliability; Evaluation of the risks linked to the dispersion of radioactive or chemical materials (impact on mankind and the environment), and to the exposure of workers and the public to radiation (this is the entire field of radiation protection); Risk management, comprizing both compliancy with the regulations relating to radiation protection and the development of decontamination processes for soils and sites contaminated following an accident. Sprinkling device Control bars Core fuel (cladding: 1 st barrier) Pressurizer Containment (3 rd barrier) Sand filter Steam generator Primary circuit (2 nd barrier) Vessel (2 nd barrier) Pump The three safety functions for reactors Control of the chain reaction Evacuation at any moment of the energy produced in the core, production which continues at the level of a few % after stopping the chain reaction (we then speak of residual power). Containment of radioactivity, the main part of this relating to the fission products formed in the fuel. Fig. 29. The three containment barriers of a PWR. Increasingly, this approach is completed by a probabilistic safety assessment (PSA), which aims to evaluate the probability of the barriers destruction, the associated radioactive waste and the consequences on the surrounding population. Here we come up against the difficulty of assessing the probability of extremely rare events. Thanks to the probabilistic safety studies carried out in the years which followed the Three Mile Island (United States) accident, operators have made provisions having effectively reduced the probability of a core melting* accident by factor 10 to 100. Nuclear energy of the future: what research for which objectives? 29

28 Another point of view is the acknowledgment of the human factor as a progress point of safety, and this from the design stage up to the dismantling, cleanup and waste management phases of nuclear installations. The analysis of significant incidents and accidents shows, in effect, that a significant part of errors likely to have an impact on the safety of the installations is linked to activities other than the conduct in control rooms (maintenance, tests, loading operations). INES gravity scale 7. Major Accidents (Chernobyl type ) 6. Accidents having limited consequences around the site 5. Accidents presenting risks outside of the site (T.M.I. type) 4. Accidents on the installations (Tokaï Mura type) 3. Incidents affecting safety 2. Incidents likely to develop later 1. Operating anomaly Fig. 30. The INES gravity scale* of nuclear incidents or accidents. Between 1995 and 2005, the French electro-nuclear fleet was the subject of thousands of level 1 incident declarations, of approximately forty level 2 incidents, and no higher level incident or accident. Risks linked to nuclear proliferation The issue consists of developing civil uses of nuclear energy, without thereby lessening the world s safety by enabling certain countries or organisations to equip themselves more easily with nuclear weapons 7. The corresponding risks must be examined from two angles The technical means that can be used to acquire fissile material needed for the construction of a bomb The easiest means of acquiring the necessary material for constructing a bomb is the enrichment of uranium. A shortcut consists of recovering highly enriched uranium used in fuels for research reactors: this is why the Americans have decided to place an embargo on fuel enriched by more than 20%, a rule generally applied today (there are however a few exceptions). A more difficult means is that implemented by the United Kingdom and France in the 1950s: to produce plutonium in reactors burning natural uranium at very low irradiation levels, enabling military quality plutonium to be produced. The extraction of plutonium requires complex reprocessing installations. Large power reactors using enriched uranium are poorly suited to the production of military plutonium, because it would be necessary to very greatly limit the irradiation of the fuel and to reprocess it. This would not be impossible, but this would be a large-scale, very expensive and difficult to hide operation. The technical means and control policies Feedback constitutes a major element in the progression of the nuclear installation safety culture. The systematic recording and analysis of incidents, and all the more so, of accidents (Three Mile Island, Chernobyl, Tokaï-Mura 5 ) must enable the operation and the safety to be improved. But serious accidents are generally and fortunately few, which does not enable them to be analyzed reliably by the existing mathematical tools. It is therefore important to pay attention to incidents as well as to near incidents, defined as events which could have led to an accident, or even to events in which human intervention has enabled a potential incident to be caught in time. Equally nuclear operators have undertaken to exchange their best practices and to inform each other regarding any significant incident, within the World Association of Nuclear Operators (WANO); the safety authorities of various countries have also established close relationships which were lacking prior to the Chernobyl disaster; and the IAEA 6 has had adopted by all of the nuclear countries a set of common safety rules and principles. 5. Three Mile Island (USA), the 28 th of March 1979; Chernobyl (ex-ussr), the 26 th of April 1986; Tokaï Mura (Japan), the 30 th of September 1999 (Note of the Editor). 6. International Agency for Atomic Energy. The keystone of the battle against nuclear proliferation is the Non-Proliferation Treaty (NPT), signed by most countries (but not all). The signatory countries commit to accepting control by the International Atomic Energy Agency of their nuclear installations and fissile materials in their possession (only 5 permanent members of the UN Security Council 8, who already had nuclear weapons in 1968, maintain the right to not submit their military programmes to international control). The controls led by the IAEA are without a doubt difficult when they concern small installations such as small uranium enrichment units. On the other hand, they are effective for large fuel reprocessing installations and power reactors. More worrying is the case of countries who have not signed the NPT or who decide to leave it. But the corresponding risks of proliferation are not directly linked to the civilian use of nuclear energy. The few countries which have developed their own nuclear weapons have moreover done it via specific means, and not by hijacking civilian installations. 7. Only the risks of nuclear weapon proliferation are mentioned here. Dirty bombs, associating radionuclides with a chemical explosive, present very real risks, but it would be infinitely easier for terrorists to seize industrial or medical radioactive sources, currently used and less protected than fissile materials from the nuclear industry. 8. Representing the United Kingdom, United States of America, France, China and Russia (Note of the Editor). 30 Nuclear safety and security

29 Risks linked to terrorist attacks This subject is frequently talked about, more particularly since the 11 September 2001 attacks. Generally, nuclear installations figure among the international installations which are the best protected against terrorist risks, due to both their massive and stocky character, and the in-depth defence provisions mentioned above. Risks linked to the transportation of nuclear materials Nuclear materials are subjected to many transformations: conversion, enrichment, manufacturing of fuel, irradiation in reactors, reprocessing, etc., all operations generally carried out in different locations, which require a great deal of transportation*, by rail, road, air or sea. Approximately 300,000 radioactive material packages are transported each year in France for the requirements of industry, the medical sector and scientific research, which represents less than 2% of all dangerous material packages transported. The tonnages to be transported are low, which does not exempt us from taking precautions in order to limit the risk of dissemination of radioactivity during these transportations. There are four types of risks linked to transportation: irradiation, contamination, criticality concerning the protection of people, property and the environment, theft or hijacking concerning the safety of materials. An important part of the precautions concerns the robustness of the containers. No major accident due to transportation is to be deplored from the beginning of the nuclear era. Risks linked to the disposal of nuclear waste Attention has been focused for a decade on the risks linked to storage (temporary by definition) and to the disposal (permanent or reversible) of high level and (or) long lived nuclear waste. Technically, four periods are to be considered: For several decades (a century in the case of the storage of MOX spent fuel), nuclear waste is characterized by a very high level of radioactivity originating from both relatively short lived fission products and actinides (curium in glass, curium and plutonium 241 in spent fuel). In parallel, there is a heat release which requires cooling; here we are in the field of industrial techniques currently used in the storage of high level waste. The beginning of disposal coincides with the beginning of the second period, when it is no longer necessary to cool the waste. The radioactivity of fission products decreases to a low value, and it is the actinides present (neptunium, americium and curium in the glass, neptunium, americium, curium and plutonium in the spent fuel) which release the heat. This release determines the dimensions, both of the waste packages (the thermal load of each package being limited) and of the disposal site (the thermal load per unit of surface area being limited). In the third period, which lasts several tens of thousands of years, even 100,000 years, the main part of the radiotoxicity inventory of waste comes from minor actinides* (for the main part, in glass, neptunium and, at the beginning of this period, americium and curium) and plutonium when the latter is incorporated into it (case of spent fuel). The potential radiotoxicity* of waste only becomes lower than that of the original uranium ore towards the end of this period. In between, the safety of a geological disposal site must be assured above all via the containment of the waste placed in disposal containers, themselves surrounded by engineered barriers; the geological barrier only intervenes in the case of failure of the latter. The work in progress on the containers, the engineered barriers and in underground laboratories aims to validate the safety analyses relating to this period. Beyond 100,000 to 200,000 years, the safety analysis considers that close containment is lost and that it is therefore the geological barrier which plays the main role of protection. A large number of behaviour models over a very long term of these radionuclides, carried out in various countries and compared in international programmes, have concluded that the doses received by man would amount to low fractions of those attributable to natural radioactivity. During the first two periods, the problems posed will mainly be national. They concern the security and safety of the storage or the passive safety of the disposal. Over a longer term, the accumulation of radionuclides in the disposal sites, however, should be considered as a legacy to the planet s future generations 9. The waste-related environmental risk may be considerable if it is badly managed (example of certain former Soviet sites such as Chelyabinsk). If they are well-managed, the impact of the waste will probably be minimal, local and delayed. No demonstration of safety could ever be provided directly, due to the time scales at stake. The role of science must probably be a little more modest: build 9. Final nuclear waste stored in well selected locations and carbon gas emissions should not be placed on the same level. For nuclear waste the risks will remain local because they would only concern at any moment the geographical neighbourhood of the disposal site, therefore in any case only a limited number of people, whereas carbon gas emissions are not controlled and their effects concern the planet s overall climate. Nuclear energy of the future: what research for which objectives? 31

30 confidence, via a corroborating stream of indications showing that all of the mishaps likely to affect the disposal have been envisaged including their consequences, etc. In short, that the latter is a robust and controlled design. Building confidence Controlling risks is not only technical and scientific: it also has a strong social component. The confidence building procedure must not stop once the conviction of experts is acquired. It is then necessary to pass from scientific uncertainty to the negotiation of the risk. Nuclear safety calls on political decisions made democratically, that is with the opinion of citizens, whose intellectual logic is different from that of scientists. Scientists and citizens have a lot to say to each other!

31 Energy in the world Since the discovery of fire, human development has been accompanied by increased energy consumption. Still today the level of energy consumption in general, electricity consumption in particular, are indicators fairly rough of development. have to compete with synthetic fuels manufactured from coal, the reserves of which are still significant. The use of hydrogen alone, without passing by a fuel cell is also a channel which must not be neglected, for land vehicles but also air or maritime vehicles. Toe / year Transport Industriy + agriculture Domestique + tertiaire Food Prehistory years Antiquity / years Fig. 31. The various uses of energy, during the ages, expressed in tonne of oil equivalent (toe) per year and per person. Energy use XVIII th century XX th century United-States XX th century The habitat represents approximately a third of the human energy consumption. This sector is likely to progress a great deal: the massive use of thermal solar energy would enable production of a large part of hot running water and residential and service sector heating. Unfortunately, the very slow rate of habitat renewal is slowing down the progress which might be carried out in this energy sector. Transportation represents a large part of the global energy consumption, and a major source of pollution. In this sector, liquid hydrocarbons seem difficult to replace in the short term, even though hybrid vehicles, combining a thermal engine and an electric engine powered by batteries, have already started to see the day fairly quickly. Hydrogen and fuel cells will reach their full potential probably in a more distant future and will Nuclear energy of the future: what research for which objectives? A particular form of energy: electricity Electricity occupies a growing part in the energy consumption of all of the developed countries, due to its privileged use in the lighting, information and communication fields, and thanks to specific advantages linked to its flexibility of use in engines and switches. Electricity is a clean energy in the phases of transportation, distribution, and end use: no pollution, and no greenhouse gases, with the exception of ozone. It is also clean in the production phase, if it is produced via nuclear, hydraulic, solar or wind power. Apart from these advantages, electricity has a significant weakness: it can practically not be stored, except in minimal quantities and at a high costs in accumulators. It can be stored indirectly (pumping stations, flywheels), but this remains marginal. Therefore it is necessary to produce it at any moment according to the immediate demand: the absolute example of tight flow! The consequence of the difficulty of storing electricity: if the network is powered by an intermittent and random source, which is the case of many renewable energies, it is necessary to plan in reserve an equivalent power source ready to replace it. The international consumption of primary energy It is fairly certain that energy consumption will increase in the next fifty years due to the increase in the world population and the increase in the standard of living in developing countries. Fossil fuels (coal, oil and gas) will still be dominant. The production of oil should, between now and around ten to twenty years, first peak then decrease. As requirements are increasing, the question will be to know how to fill this lack by more environmentally friendly energy sources. The problem is not to oppose energy sources, but to find the best use for each one of them in order to reach the most effective energy mix. 33

32 Units It is important to specify at which stage of use energy is counted. When the assessment is made at source (coal mine, oil well, hydraulic dam, etc.) we speak of primary energy, which is what we will do in the rest of this text. Useful or final energy can also be counted. Due to the yields from transformation and various losses, almost three times more primary energy is needed than useful energy, directly linked to the service sought. When we talk about the energy consumption of the country, we generally count it in tonne oil equivalent (toe). Approximately, 1 tonne of coal equals toe and 1 MWh of gas equals toe. Things become complicated when it involves expressing in toe the energy produced by a primary electricity source, which does not come from the conversion of a fossil fuel (hydraulic, nuclear, wind electricity, etc.). In 2002, France rallied to the International Energy Agency conventions: The nuclear (or geothermal) MWh equals 0.26 toe, quantity of oil that it would be necessary to burn to produce one MWh in a thermal power plant. This average of 2.3 toe/year per person hides a very large regional disparity, which reflects the great North-South division in terms of development: whereas an American consumes 8 toe every year, a European or a Japanese person makes do with 4 toe, and an Indian citizen lives with only 0.4 toe per year. Energy consumption per inhabitant (TOE) India China Emerging countries World Japan France Germany EU United States 0,2 0,7 0,8 1,6 3,9 4,0 4,1 3,8 8, Fig. 33. Can these inequalities last? Source: Etude IIASA/WEC Global Energy Perspectives, Today, the Earth s inhabitants consume on average 2.3 toe per person each year, which leads to a cumulated annual consumption of primary energy, all sources together, of 9 billion toe, 9 Gtoe/year. Gtoe / year World population in billions A B C A: Steady growth B: Average growth C: Eco-friendly growth 2100 Fig. 32. The world s population approached half a billion individuals at the beginning of the christian era. It reached a billion towards the middle of the 19 th century then, by a fantastic acceleration of demography, the current figure of 6 billion in only 150 years. The world population is now increasing at a more moderate rate, but, at the present rate, we will without a doubt reach 10 billion during this century. Distribution per source The table below provides the distribution of the world energy consumption between the various primary sources, in 2000, according to the International Energy Agency: Source Million toe % Solid fuel 2, Oil 3, Gas 2, Nuclear Hydroelectric New Renewable Energies (NRE) Total (commercial) 9, Coal Oil Gas Nuclear Hydroelectric NRE Fig. 34. The world energy consumption between the various primary sources. 34 Energy in the world

33 It can be seen that fossil fuels amount to 90% of the commercial primary energy used on the planet, and still over 80% if non-commercial energy is taken into account. The figures speak for themselves: there is no way that the increase in the contribution of new renewable energies (NRE) can alone cover the increase in requirements or replace nuclear power as wished by certain people. In any case, not in the decades to come. Even if the OECD countries made spectacular strides in increasing energy efficiency, the requirements of developing countries are such that the energy consumption could not increase less quickly than the population itself. All the more because the OECD countries and those of the former USSR have now stabilized, and the 4 billion human beings which will increase the world population during this century will originate from current developing countries. In order to face these gigantic requirements, we will not have too much of all of the energy sources that mankind knows how to master! Fossil energies The proportion of fossil energies should remain largely preponderant in future decades. It should represent, according to the International Energy Agency, 90% of the commercial energy supply by 2030, hydrocarbons (oil and gas) representing approximately 65%. Oil The proven reserves of oil currently represent approximately around forty years of production at its current rate of consumption. Burgan Zuana Ghawar Annual discover Smooth on 5 years Production The rate of discovering new exploitable oil fields has decreased since the 60s, which implies a rapid exhaustion of conventional resources if the current rate of consumption is maintained. Thanks to new discoveries and also a better recovery of oil on site, it should be possible to exploit much larger resources.the latter will nevertheless be much more expensive than those recovered today. By lowering the costs, technical progress should enable the development of deep sea production and the exploitation of very deep deposits. Beyond that, the resources in non-conventional oil, and in particular, the extra heavy crudes, asphaltic sands and kerogen shale are considerable. Natural gas The proportion of natural gas in the world s energy inventory continues to increase, given its advantages: lower impact on the environment than coal or oil (no dust, better efficiency for the generation of electricity with combined cycles, turbines, etc.), its flexibility of use, the importance of its reserves greater than that of oil (it currently represents more than 60 years of consumption at the current rate). There are also considerable reserves of methane hydrates (without a doubt more than double the quantities of fossil fuels which are yet to be exploited) that are trapped at the bottom of the sea or permanently in frozen ground. We still however do not know how to recover the latter technically. There are also uncertainties regarding the energy efficiency and the economic cost of this recovery. Coal Coal, after a period of decline, may return in force, given the importance of its reserves, which represent several centuries of consumption at the current rate, in particular by implementing gasification systems, which enable it to be used in a cleaner way. A good illustration of this is the Futuregen project, which was started in the United States of America in order to lead to the industrial demonstration of a system of electricity production from clean coal with sequestering of CO 2. Gach Saran Tia Juana Fig. 35. The rate of discovery of new exploitable oil fields has decreased since the 60s, which implies a rapid exhaustion of conventional resources if the current rate of consumption is maintained. Nuclear energy of the future: what research for which objectives? 35

34 Fossil energies, CO 2 and climate change In the future, we must be able to answer to two major questions: how to best manage the finite reserves of fossil fuels and how, moreover, to respond to the risks of climate change, by restricting greenhouse gas emissions.the increase in the content of these gases in the atmosphere is due, without doubt, to human activities, and mainly to the use of fossil fuels, coal, oil and gas. If the phenomenon is not controlled very quickly, the global climate of the planet will be affected for a long time, with potentially catastrophic effects. Past and future concentrations of CO 2 in the atmosphere Projections Direct measurements Ice core data ppm The relative contribution of the various sources of electricity to the production of Greenhouse Gases (GHG) Among the various figures of the literature, here are the results from the LCA (Life Cycle Analysis, ISO standard) carried out by EDF, in CO 2 gram equivalent per electric kilowatt hour: System Operation Remaining Total life cycle g/kwhe Coal 600 MWe Fuel-oil Gas (turbine combustion) Diesel Hydraulic pumping Photovoltaic/solar Hydroelectric Nuclear Wind Fig. 36. The concentrations of CO 2 expected during the 21 st century are two to four times that of the pre-industrial era. The sequestering of CO 2 is an actively pursued avenue of research. The options which seem to be the most interesting are those which consist of storing CO 2 in exhausted hydrocarbon deposits or in deep aquifers 10.Work is yet to be carried out in order to reduce the costs, which are currently in the order of per tonne of CO 2 avoided, and to ensure the longterm security and longevity of the disposal site. Various demonstration projects are in progress or planned (Sleipner, Weyburn, In Salah, etc. deposit). Importance R&D programmes have also been undertaken, in particular on the European level. However the capture and disposal of CO 2 may only provide a partial response to the problem posed: indeed, this solution can only currently be envisaged in large scale static systems, which excludes it from the transportation and habitat sectors which represent a very large part of the emissions. A possible way out would be a large-scale use of hydrogen produced without CO 2 emission, but this cannot be envisaged in the short-term. 10. Permeable rocks with a high water content (Note of the Editor) As seen, it is not entirely true to say that nuclear, hydraulic or wind power does not produce any greenhouse gases, because the construction of power plants, dams or windmills requires concrete and steel, the production of which itself releases GHGs. But their contribution remains truly marginal! Renewable energies The history of humanity is dominated by the use of renewable energy, because the latter started to be used when humans discovered fire, approximately 500,000 years ago. Renewable resources are immense; the most abundant, solar radiation, represents kwh annually, that is to say more than 5,000 times the entire global consumption of primary energy. But these resources are generally intermittent and many require disposal in order to respond to the demand of modern societies, where the consumer wants energy when and where he needs it and not when it is available. The biomass and the accumulation of water in reservoirs represents an energy storage which improves availability, and geothermy diffuses a heat flux fairly continuously. In comparison to modern resources, often highly concentrated, renewable resources have the disadvantage of a low density; therefore they must rather be transformed there where nature delivers them. Finally, even though the resource is free, the cost of many renewables is still too high in relation to other energy sources.this is mainly due to the additional investment costs for conversion systems. These additional costs resulting either from the too low density of energy, or from the market which is still too little developed, or from the technology which 36 Energy in the world

35 has not yet reached its asymptote of minimal cost. The limitation of non-renewable resources and their impact on the environment results in renewable energies experiencing increased interest. But, at this moment in time, hydraulics for producing electricity, and biomass for producing heat, easily dominate the renewable energies market, thanks to their competitive cost. They alone represent almost 20% of the primary resources exploited. The short and medium-term avenues are energy savings, the replacement of coal by natural gas, the development of renewable energies, and, last but not least, that of nuclear power. The field of energy is a coveted field, the country which will find and develop good technologies will reap an enormous competitive edge. It is therefore important that Europe be in the race for the highest international level of innovation. A European political intent has led to supporting the mechanism of restricting greenhouse gas emissions; the Kyoto protocol is the first step towards this. One of the aspects of this intent is secured by the European directive of the 27 Electricity production in Europe (in 2000) per country and per source September The latter specifies that the generation of electricity from renewable sources must pass from Billions of Kwh (gross) Billions of Kwh (gross) % to 22% in Europe in 2010, from 15% to 21% in France The energy challenge The 20 th century has bequeathed us a double challenge in the field of energy: To confront the energy requirements of a world population multiplied by 10 between 1750 and 2050, consuming 10 times more per inhabitant; To control local, regional and planetary pollution (climate). It has left us the means, but mobilisation must be complete in order to rise to these challenges Nuclear D B E F U-K NL I S Total European Union Hydroelectric Wind Coal Gas Fuel Fig. 37. An illustration of the disparity of electric energy sources in various European countries. Although the increase in requirements and the aggravation of risks originates from poor countries, rich countries can and must provide their contribution: controlling energy, technology transfer, reduction of their GHG emissions. The answers to be provided are not necessarily the same in every country because priorities depend on the state of development, the domestic resources, financial capabilities and the cultural context, but the challenge is to be met on a global scale because the effects on the environment are global. Denmark United-States Germany UK Japan Sweden France Fig. 38. The choice of energies: CO2 emissions per kwh throughout the world (gc/kwh). Nuclear energy of the future: what research for which objectives? 37

36

37 The economy of nuclear power The cost per kwh produced by high power water nuclear reactors has been the subject of numerous studies, the most complete of which are probably those conducted by the French Ministry of Industry, which is based on a very important and well-known programme, and that carried out by the Professor Tarjanne for the Finnish government and which supported its recent decision to build a new reactor. The DGEMP study notes that all of the nuclear power costs are in fact taken into account, contrary to that which occurs for other energies: in particular the provisions for waste management and for the dismantling of installations. This overall cost is evaluated at approximately 30/MWh, of the same order of magnitude as the internal cost of the kwh in a combined cycle gas power plant. The Finnish study compares the costs of the kwh of a 1,250 MW nuclear power plant and those of fossil fuel and wind power plants, with the conventional hypotheses regarding the lifetime of the power plants (40 years) and the availability of the installations (90% for thermal power plants, 2,200 Cost of electricity generation Euro / MWh Fuel Fig. 39. Cost of electricity ( /MWh) for various sources of primary energy, compared to the ELSPOT price, the Scandinavian kilowatthour market. The Finnish study from March 2002 concludes that nuclear power is the most economical energy source when power plants operate more than 6,000 hours per year. Only the internal costs have been taken into account. A possible eco-tax on carbon would further improve the competitiveness of nuclear power ELSPOT ELSPOT Nuclear Coal Gas Peat Wood Wind Price Price max 2000 maxi Price in November Interest rate = 5.0 % Maintenance and operation Generation cost excluding subsidies and tax benefits Cost of the capital Nuclear energy of the future: what research for which objectives? man/year for wind turbines). As its French counterpart, the Finnish study gives the advantage to nuclear power. Between 1974 and 1985, in France in particular, nuclear power enjoyed a comfortable margin of competitiveness. Then, the oil counter-shock very rapidly brought the price of fossil energies back down to the level that they were prior to 1974, and this situation lasted until 1998, profoundly changing outlook. On the other hand, gas turbines, benefiting from technological effects from the aeronautical industry, achieved spectacular progress in efficiency, unit size and thus price, whereas the strengthening of safety and the increase in regulations has somewhat increased the nuclear investment. Since 2003, the scale has once again tipped in favour of nuclear power and this trend should last and be accentuated with the inexorable rise in the price of hydrocarbons, accompanied by the progressive growing scarcity of resources. The evaluation of the competitiveness of nuclear energy must be made over a long time period, comparable to the lifetime of nuclear plants. The comparison depends on the price performance of other primary energies during this time period. For hydrocarbons, this evaluation is very uncertain. What is certain is that the nuclear power costs are stable and predictable. The price of uranium raw material only intervenes for a very small part in the price of the nuclear kwh.this favorable situation protects nuclear energy from fluctuations in the raw materials market. Moreover, 90% of the expenditure takes place on the national territory, with the location corresponding to the use, and with favorable consequences on the balance of payments. Today, existing nuclear power plants, partly amortized, constitute important sources of profit. On the other hand, the initial investment necessary for constructing new ones is high and difficult to assemble. It will probably be necessary to resort to new financing structures in order to finance such heavy investments in a largely deregulated economy focussed on the short-term. Evidence of feasibility is starting to emerge: for example, a consortium of Finnish paper manufacturers is financing one part of the new PWR reactor recently ordered in this country. These credit arrangements must take account of the fact that the heavy investments of the nuclear power industry, combined long-term with the 39

38 return on investment, makes the profitability of nuclear power particularly sensitive to the interest rate for necessary loans. Nuclear Gas combined Wind cycle Real interest rate 5% Real interest rate 8% Real interest rate 5% and carbon tax 20/t Fig. 40. The results of the Finnish study compare the price in of the electric Megawatt-hour for various production modes. Two real interest rates are taken into consideration: 5 and 8%. The competitiveness of nuclear power is indisputable for the low capitalisation rate (5%). Nuclear power remains competitive compared to gas up to a real interest rate of 8%, which leaves a fairly comfortable margin. /MWh Coal Oil Gas Bio Nuclear Solar Wind Fig. 41. Health and environmental costs, called external costs for various energy sources Source: ExternE, J. Weisse, March In normal operation, energy production systems impact our environment and our health, which should be taken into account if we want to compare them. For some activities, it concerns liquid or gaseous waste, for others it is a noise disturbance or simply the degradation of a tourist site. They also involve possible accidents, the consequences of which must be taken into account. The ExternE study, carried out in collaboration between the European Commission and the US Department of Energy, aims to identify and even quantify external costs and profits, that is the positive or negative effects of various energy systems, not taken into account in the direct economic assessment. It emerges from these studies, carried out within a European framework, that the external costs of nuclear energy are particularly low. We will not enter into the debate here regarding the value allocated to plutonium, which may be considered as waste or as a precious resource, according to the recycling policy chosen. Recent studies suggest that the economy of fissile material permitted by the recycling of plutonium barely compensates the costs associated with recycling. The elements of choice are furthermore not only economical but also, and above all, political, because considerations regarding the radiotoxicity of final waste, the possible continuation of the nuclear programme with fast neutron reactors or the proliferation of nuclear materials influences strategic choices. 40 The economy of nuclear power

39 Nuclear power throughout the world Several events have recently reminded us of the advantages of nuclear power and open new prospects to the sector Firstly, there is the European Commission Green Book 11, which recognizes in nuclear power a competitive energy source capable of responding to the double issue of the safety of the energy supply and the reduction of greenhouse gases. In the same way as controlling the energy demand and developing renewable energies, nuclear power is currently recognized as an essential component of a more balanced European energy mix, privileging energies that do not emit greenhouse gases. An equivalent document, the Report of the National Energy Policy Group, published in the United States in May 2001, delivered a similar message. Awareness of the sharp increase of the demand for primary energy throughout the world leads to the recognition that all energy sources will be necessary in order to match the needs, including nuclear power, which produces practically no greenhouse gases. Most countries thus currently integrate nuclear power into their reflection on the short and medium term (up to 2020) and long term (2020 and beyond) energy policy. This is actually the case of the United States: between new requirements and the replacement of aging installations, the American administration thus evaluates the number of electric power plants between 1,300 and 1,900 (that is power on the order of 400 GWe) which must be installed between now and 2020, all sources together. The availability of American nuclear power plants has spectacularly improved, which constitutes the main reason for the renewal of their appeal in the United States. A large number of the 104 American reactors have obtained from the safety authority the extension of their operation beyond the initial expected duration, and these second-hand reactors are resold between electricians at the price of new reactors. Even if all of the conditions have not been met for significant investments in new nuclear power plants over the very short term, the use of nuclear power for the medium-term remains unavoidable in order to satisfy part of the high demand. Several tens of GWe of nuclear origin will without doubt be necessary by this time Fig. 42. Nuclear power plants operating worldwide It is in this context of vulnerability of hydrocarbon supply and environmental constraints, that the National Energy Policy report, submitted to the President of the United States in May 2001, concludes with the need to resume the development of nuclear power in this country. At the same time, the Nuclear Power 2010 initiative was launched in order to accelerate the process of granting authorisations in view of the deployment of advanced reactors from 2010.The decision-making process regarding the opening of the Yucca Mountain nuclear waste disposal sites has been initiated with the positive vote from Congress in 2002 regarding this project, which had already obtained President Bush s support.the Department of Energy (DOE) is moreover very active within the scope of the International Forum on the fourth generation. The importance of closed fuel cycles ( Advanced Fuel Cycle Initiative ) is also being largely reconsidered, beyond the past position of opposition to any reprocessing. Finally, it is important to point out the propositions of research programmes from several national laboratories aiming to relaunch R&D and the related budgets. 11. Published in November Nuclear energy of the future: what research for which objectives? 41

40 USA Power Plants by 2020 including nuclear (> 50 GWe) FINLAND 5 th reactor FRANCE new EPR reactor KOREA nuclear capacity increase + 9 GWe by ~ 2015 CHINA nuclear capacity increase > 30 GWe by 2020 JAPAN nuclear capacity increase + 21 GWe by 2012 INDIA nuclear capacity increase from 2.5 to 20 GWe by 2020 Fig. 43. There are currently 34 nuclear reactors being built throughout the world, and nearly as many projected. Russia, in spite of its economic difficulties, seeks to participate in the global reflection on nuclear power and has taken several initiatives in this direction, in particular the launch of a global concertation, with the IAEA, regarding nuclear power of the future (INPRO exercise) and the vote for a law on accepting foreign nuclear waste on Russian territory enabling nuclear fuels to be offered on a lease basis. Re-establishing strong economic growth, Russia is now showing a willingness to succeed with its civilian nuclear development programme with the completion, then the commercial commissioning of power plants, the construction of which was stopped following the Chernobyl accident in The country is also very active with its fast neutron reactor development programme. The United States and Russia are moreover making specific efforts, within the framework of nuclear disarmament, to convert and use fissile materials of military origin. This has created a common reflection, with a high involvement from France, regarding the cycle of these materials (including in the United States) and on reactors best suited to reach this objective. In Asia, China, whose GDP has seen near 10% annual growth over the last few years, estimates the needs for new electrical capacities at approximately 20 GWe per year during the next 20 years.this huge figure indicates how important it is for this country to increase its production capacities. In the 1980s, China launched itself, up to now within the framework of a privileged relationship with France, into a nuclear equipment policy with a willingness to control all of the technologies associated with the construction of reactors. Even though nuclear power only represents 1.5% of its capacity with 8 reactors operating commercially, China projects having by 2020 a capacity in the order of 35 GWe of nuclear origin, that is the equivalent of 20 to 30 new reactors. The nuclear proportion could therefore reach 4 to 5% of the capacity, thermal and hydraulic power remaining largely in the majority. The recent soaring oil prices and the awareness of the energy dependence regarding exterior supplies nevertheless leaves open the possibility of an acceleration of the development of the Chinese nuclear programme. Japan, which owns little natural energetic resources, has adopted a strategy similar to France s in the 70-80s. It already has 53 reactors which generate 45 GWe, ie approximately 34% of the national electricity. Four reactors are in the process of being built and around ten additional units are active projects. The next report from the Japanese Atomic Energy Commission will update the deployment projections of the electro-nuclear fleet by The previous report took into account a need in the order of 20 new reactors between now and Japan, whose population is declining, is nevertheless currently beset by political and institutional difficulties regarding nuclear power: the latter is struggling to regain public confidence after numerous matters which have punctuated the life of the sector during the last few years.the country may therefore downsize its ambitious programme. 42 Nuclear power throughout the world

41 In South Korea, 19 nuclear reactors represent approximately 38% of the national electricity generation. This country is currently constructing 2 nuclear reactors and is planning to increase its capacities by constructing an additional 8 reactors in the next 12 years. In the longer term, Korea, which is poor in energy resources, plans to double its capacity installed in India, with its one billion inhabitants and in spite of its low per capita consumption (0.5 toe/inhab/year), already figures among the largest energy consumers and is faced with important energy deficits. Fourteen nuclear reactors, of low power and mainly CANDU* technology, are currently in operation and the Indian government hopes to increase the nuclear capacity of the country for it to pass from approximately 3 GWe today to 20 GWe between now and In order to do this, India intends to increase its production capacity partly from reactors developed locally and partly by turning to foreign partners in order to have access to light water reactors*. It should be noted that the country is actively pursuing its fast system and thorium system development programme, given its national reserves. Brazil relies heavily on its hydroelectricity but has already commissioned two PWRs, the first ordered from Westinghouse and the second from Siemens. Framatome- ANP is expecting the decision to complete the construction of the third reactor ordered, at the time, from Siemens. The Brazilian Ministry of Science and Technology has furthermore declared itself in favour of the development by its country of research on nuclear technology. South Africa, with 2 powerful reactors in operation, is developing a 100 MWe high temperature reactor, of the Pebble Bed Modular Reactor (PBMR) type, in partnership with BNFL 12 (GB) and a future partner yet to be defined. This concept of small reactor, based on the German pebble technology and cooled with helium, mainly targets a faster return on investment than that of the PWRs and would also present the interest of being accessible to small countries, given the smaller initial investment. The ESKOM consortium announced in 2003 that it was now ready to pass to the development and construction of a PBMR demonstration reactor. And in Europe? Each European country is sovereign in the choice of its energy options, which leads to a panorama of very different options for nuclear power. Some countries make great use of nuclear power (nearly 80% of the French electricity generation is of nuclear origin; others have none at all (Ireland, Austria, Norway, Denmark, Italy). 12. British Nuclear Fuels (Note of the Editor). Nuclear energy of the future: what research for which objectives? Italy and Austria have declared themselves against nuclear energy since the 1980s; in the same decade, Sweden decided to pull out of nuclear power by 2010 but up to now has shutdown only one power plant; Germany decided to pull out of nuclear power in 2000 and Belgium in 2001; the United Kingdom deferred the decision to renew its reactors to a later date... Conversely, in May 2003 Switzerland refused to pull out of nuclear power, and Finland ordered a PWR nuclear reactor on 18 December France has just decided to construct a PWR reactor on the Flamanville site. This common absence of vision between the European Union countries creates an unfavourable global political environment for nuclear power, and yet this energy currently occupies an important place in Europe: It contributes 35% of electricity production; It represents an important industrial sector on the international scene, for the supply in reactors (in particular those of 3 rd generation such as the PWR and the SWR 1000 from Areva), for operating power plants, and for the fuel cycle; The European research on radioactive waste management, in particular that conducted in France since 1991, is among the most advanced in the world and from 2006 will enable a new management strategy to be decided regarding all of the waste produced by nuclear power plants; European community R&D programmes also dedicate significant resources to thermonuclear fusion. On 1 st May 2004, the European Union passed from 15 to 25 members: thus 5 nuclear countries out of 10 joined Europe with 23 reactors in commercial service. The importance of these countries integration extends to the nuclear energy sector, in particular in terms of safety and waste management. The prospects of collaboration, in particular with the Czech Republic, Slovenia and Hungary are numerous. Signs of renewed interest are appearing: The European Commission Green Book, published in 2001 concludes, of course in very cautious terms, on the need to reconsider the nuclear option in order to face energy supply problems and to respect the Kyoto commitments; Through private investors, Finland confirmed its choice of construction of a fifth reactor by placing an order at the end of 2003 for a PWR reactor with the Areva group; Sweden, after closing Barsebäck 1 at the end of 1999, postponed sine die the shutdown of its nuclear plants because they could only be replaced by a Danish electricity import (with coal) with the consequence, well perceived by public opinion, of acid rain. Opinion polls are currently in favour of 43

42 pursuing the nuclear activity in this country which 20 years ago opposed it by referendum; Switzerland consulted its population by referendum in May 2003 regarding popular initiatives which should have led to eventually pulling out of nuclear power: A quite clear refusal of renouncing this energy emerged, even though in 1990 a moratorium banned the construction of any new power plant of this type; Germany concluded a political contract, well in the social consensus tradition of this country, to postpone the real choices to a later date, whilst protecting the main thing, that is the operation of existing power plants and thus the country s safe supply in electricity; Belgium more recently turned to a similar track, but changes in its political landscape could eventually lead to a revision of its position and a possible repeal of the law for pulling out of nuclear power; Great Britain is conscious that with the foreseen shutdown of its ageing Magnox reactors between now and the end of the decade and the exhaustion of North Sea pools within 25 years (in 2004 the country had just crossed its peak oil ), the electricity supply will be entirely dependent on imports in the not too distant future. A re-examination of the energy policy is in progress and a White Book came out in The latter certainly indicates a marked willingness to significantly reduce CO 2 emissions by 2020 but did not clearly define the British position regarding the nuclear option. Lithuania France Belgium Ukraine Sweden Bulgaria Slovakia Switzerland Hungary Slovenia Japan Taiwan South Korea Germany Finland Spain United-Kingdom Armenia United States Czech Canada Russia Argentina Romania South Africa Mexico Netherlands India Brazil China Pakistan Kazakhstan % Fig. 44. The nuclear proportion in global electricity generation. Coal (39 %) Finally, in France, the Parliamentary Bill regarding energy confirms the major contribution of nuclear energy in the future national energy mix and the importance of launching the first PWR reactor in order to guarantee the renewal of the current fleet in due course. EDF has entered into the process of launching the EPR by selecting the Flamanville (Manche) site for the first reactor. Fig. 45. World electricity generation. Hydroelectricity (19 %) Nuclear (16 %) Gas (15 %) Oil (10 %) In this context favourable to a rebirth of nuclear power throughout the world, the current objective is to enhance all of the European potential in the development of future nuclear systems, with a just return of the profits to come. This objective implies the different types of issues for which research laboratories and their industrial partners are preparing themselves: Research issues to develop the key technologies for sustainable nuclear power, an issue which presumes a willingness from political and industrial players to continue investing in the R&D for the nuclear power of the future; Industrial issues to enhance the experience acquired in the previous development of prototype reactors or advanced processes regarding the fuel cycle and to commercialize it; 44 Nuclear power throughout the world

43 Industrial issues also to be recorded in international consortiums called upon to market the systems of the future; Also issues of development training in international cooperation, which would lead to sharing R&D whilst researching the reduction effects of national efforts by synergies, and co-financing opportunities for large research tools or prototype installations.

44 Nuclear power: the main avenues of research To support the current nuclear industry; To provide effective and acceptable solutions to the problem of long-lived and high level waste management, and to better understand the impact of nuclear activities on humans and the environment; To design and evaluate new generations of nuclear systems (reactors and cycles). 46

45 The near future: research supporting the existing nuclear power The nuclear industry has reached its full potential. However, the margins of competitiveness can still be improved: By improving the profitability of the fleet through more efficient use of nuclear fuel; By extending the lifetime of existing reactors. The global fleet of reactors is aging rather well, and many electricians throughout the world envisage operating these existing reactors longer than the lifetime for which they were initially designed. It is still necessary to obtain authorisations, and for this, demonstrate that the aging of reactor components is foreseeable and controlled; By preparing the replacement of the fleet of current PWR reactors with evolutionary third generation reactors, endowed with an improved efficiency and a (further) increased level of safety 13. These three avenues of improvement for the near future require R&D. CEA takes on a large part, in close partnership with the French nuclear industrialists, Areva and EDF. Using nuclear fuel more efficiently Industrial issues At the time of the start up of the programme for the construction of PWR power plants for electricity production, in the 70s, one of the arguments put forward (apart from the energy independence) was linked to the relatively low cost of the fuel cycle. Indeed, the fuel cycle proportion in the cost of the kwh (30% including the upstream and downstream sections of the cycle), did not bring about particular optimisation efforts for the fuel performances. Today, given the updated economic reports between the various energy production systems, there are important productivity gains to be achieved thanks to nuclear fuel and to its management means. 13. Third generation reactors will be dealt with in the following chapter. Nuclear energy of the future: what research for which objectives? For electricians it involves increasing the overall efficiency of its nuclear fleet in order to be competitive in an open market: By increasing the burn-up rate* of fuel assemblies; By extending the irradiation campaigns; By reducing the number of assemblies on each reloading (flexibility of the reloadings); By reducing the operating constraints, in particular during transient periods imposed by the monitoring of grid loading (these transient periods in fact test the fuel, and the development of a fuel capable of resisting rapid changes in reactor speed is an important issue); By controlling the equilibrium of the fuel cycle over the entire fleet, a policy of matching the reprocessing recycling flux. R&D objectives and challenges regarding PWR fuel The maximum irradiation (average per assembly) is currently 52 GWd/t whereas it was 33 GWd/t in the 80s. This important increase was obtained mainly thanks to: A better knowledge (associated with comprehension and modelling) of the behaviour of the fuel in irradiation provided by the R&D and the feedback from standard or experimental fuel irradiated in PWR cores, enabling optimised dimensioning; Progress on fuels themselves (cladding material, pellet, importance of the fuel ceramic microstructure). At present, the burnup rate is restricted by the strength of the cladding (the fuel is removed from the reactor before the cladding breaks, or rather, before it risks breaking in incidental situations). With the objective of reaching burnup rates exceeding 70 GWd/t within the next decade, a certain number of developments and/or confirmations are necessary. These developments concern numerous, often combined, phenomena (corrosion of the cladding, internal pressure, mechanical behaviour of the assembly and rods in incidental and accidental situations, etc.). 47

46 CEA, in close collaboration with its industrial partners, has set up R&D programmes regarding fuel, based on its experimental means and on its expert capacity. R&D programmes regarding fuel Responding to industrial needs In the short and medium-term, the R&D needs expressed by industrialists require a follow-up or even an increase in R&D efforts in the following fields: The behaviour and reliability of the mechanical structures of fuel assemblies for high burnup rates. The progress targeted consists of a reduction of the mechanical wear and tear of rods thanks to a better control of their vibratory behaviour in the reactor core. They undergo particular tests carried out in representative situations (temperature conditions, pressure, chemistry and geometry of the reactor cores). These tests, carried out on the CEA/Cadarache Hermès installation serve to validate the modelling and simulation of the behaviour of assemblies, and to demonstrate that the main phenomena at stake are understood and controlled; One of the R&D objectives in particular regarding MOX fuel is to increase its competitiveness by increasing its burn-up rate. The aim is to produce a ceramic capable of effectively retaining fission gases. It has been recently demonstrated that the use of additives introduced into the oxide powder prior to sintering enables the homogeneity to be improved and the size of the surrounding grains to be significantly increased, two important conditions for minimising its gaseous release under irradiation. The current irradiation experiments in progress on these new ceramics will enable the gain obtained in burnup rate to be quantified. The experiments mainly consist of instrumented irradiations (for an example in the Osiris reactor in Saclay), followed or not by thermal annealing associated to measurements of the fission gas release (in hot laboratories, for example LECA/STAR in Cadarache). Post experimental examinations use the following conventional tools: electron microscopy, microprobe, mass spectrometry of secondary ions, with the particularity that the corresponding devices are adapted for the examination of highly radioactive objects; In the field of cladding, even with current materials or in process of deployment (such as the zirconium-niobium M5 alloy), the behaviour of the claddings in more demanding conditions (high temperature oxidation with steam, hydridation, fragilisation, etc.) must be explored further particularly for the safety demonstrations of new modes of fuel management; MIC02F1 Addition of Cr 2 O 3 before dilution MIC0F1 MIC03F1 Addition of Cr 2 O 3 during dilution Fig. 46. Microstructural analysis of advanced MOX fuel ceramics.the photo opposite compares the local content in plutonium of the MOX pellets sintered respectively with and without chromium additive. The second are much more homogenous. 48 The near future: research supporting the existing nuclear power

47 By 2010, the qualification of a fuel much less sensitive to the pellet-cladding interaction is a main objective of the EPR project, in particular in order to improve the reactor s performances, simplify its design and to minimise the constraints linked to the monitoring of the power grid loading. Burn-up rate 70,000 MWd/t HTC (900 / 1,300) 62,000 MWd/t Galice (1,300) Oxide thickness (µm) ,000 MWd/t Cyclades (900) Gemmes (1,300) 47,000 MWd/t Alcade (N4) Garance (900) 39,000 MWd/t Zy Year Burn-up Fig. 47. Example of current progress regarding the fuel cladding: the zirconium alloys used for the cladding are subjected, in the presence of water, to corrosion which tends to spiral out of control as the oxide layer grows, which limits the time that the fuel stays in the reactor and the temperature of the fuel and coolant. Recent progress in the composition of cladding alloys enables corrosion to be considerably reduced and these limitations to be postponed. M5 Fig. 48. Changes in core management involves research on the behaviour of fuel with high burn-up rates. Progress carried out on the strength of the cladding enables much higher burn-up rates and a better use of the fuel to be envisaged. Thanks to this type of progress and to this small steps policy, the burn-up rate of nuclear fuel passed from 39 GWd/t to 52 GWd/t in ten years and progress is still going on: 70 GWd/t is targeted in In order to carry out these studies correctly, CEA possesses heavy facilities: the LECA and LEFCA laboratories enable experimental fuel elements to be manufactured; the Osiris reactor (Saclay) enables the irradiation; the PELECI (Saclay), LECA (Cadarache) and ATALANTE* (Marcoule) hot cells enable these irradiated elements to be analyzed. Some of these heavy facilities are recent, others are aging. This is the case of the Osiris reactor, which must be replaced by 2014 with a powerful and multipurpose research reactor intended to cover most of the European experimental irradiation needs: the Jules Horowitz reactor*. Fig. 49 and 50. The Osiris reactor. Experiments on fuel mainly consist of irradiations in experimental reactors such as Osiris. These are long experiments, they are intended to validate the modelling, and to provide confidence in its predictive capabilities. They also serve to qualify advanced fuels, prior to their use on an industrial scale. Nuclear energy of the future: what research for which objectives? 49

48 Ventilation chimney Electronic-control building Orisis reactor Hot cells hall Isis reactor Crown building Offices Isis laboratories Cooling towers Waste containers Fig. 50. Another objective for CEA is to upgrade R&D methods, in particular in the sectors where heavy experimentation is widely used. This involves either taking best advantage of the entire experimentation or substituting it, where possible, with a more analytical experimentation based on a more cognitive approach of phenomena and sizes which govern them. This must be done in complementarity with modelling development. Fig. 51 et 52. The future Jules Horowitz reactor should diverge in Cadarache in Fig. 53. The Atalante facility, in Marcoule. 50 The near future: research supporting the existing nuclear power

49 Developing modelling and simulation CEA has modelling tools and continues to develop calculation codes, continuously improving the predictions on the behaviour of reactors and of their components. Apart from the Cathare thermal-hydraulic* code, the future EDF/CEA Descartes (neutronics*), Neptune (thermal-hydraulic) and Pléiades (fuel) platforms must be mentioned. The fuel modelling effort consists of extending the field of validity of fuel behaviour models and to assure their qualification by a specific experimentation.this field in particular covers the modelling of the thermomechanical behaviour of the rod, the description of the microscopic mechanisms of the fission gas release on the microstructure level, the pellet-cladding interaction and the irradiation-induced swelling. All of these models are introduced in the Pléiades software application codeveloped by CEA and its industrial partners. Extending the lifetime of existing reactors Planned initially to last for approximately forty years, nuclear power plants age rather well, as the feedback on the global fleet shows.that said, nuclear power plants see their profitability increase considerably once the initial investment is amortized. The extension of the lifetime of reactors is therefore a major issue for electricians. This is why many nuclear operators throughout the world are currently requesting that their country s Safety Authority authorize extension of their facilities lifetime. The French fleet of reactors is younger than the world average, but EDF also wishes to extend the lifetime of its reactors. It is still necessary to demonstrate that the system s safety is preserved. The extension of nuclear power plants lifetime requires a very good knowledge and very good mastery of the aging mechanisms of all of their components. It is also necessary to have reliable diagnosis and control means. CEA carries out research in these two fields. The aging mechanism of a nuclear reactor s components are very diverse. Some such as material fatigue, corrosion under stress, corrosion-erosion, and wear and tear by friction are absolutely conventional and are found in many other installations or industrial objects. Other mechanisms are more nuclear specific, in particular the fragilisation and swelling of steels due to irradiation and corrosion due to radiation.the various mechanisms do not act in isolation: it is their combined action which contributes to accelerating the aging of a nuclear power plant s components, and which is to be controlled. The aging of the power plant s components Fig. 54. Modelling of the pellet-clad interaction. Subject to irradiation, the fuel ceramic pellet tends to swell, due to disorders of the crystal lattice caused by the irradiation, and due to interstitial atoms generated by the nuclear reactions (in particular fission gases). This swelling, combined with pressures inside and outside of the zirconium alloy cladding, places the latter under stress. It is important to correctly model these stresses and their evolution in time, in order to control the risk of breaking the cladding and releasing radioactivity into the primary circuit of the reactor. The vessel* of the primary circuit of water reactors is one of the elements presumed to being non-replaceable. It constitutes part of the second containment barrier: its mechanical strength must be kept, even in accidental conditions. It is also the subject of a specific lifetime monitoring and evaluation programme.thus, on each ten-year visit, EDF presents the Safety Authority with a vessel maintenance file justifying its ability to fulfill this safety function for the next ten years. The main phenomenon of vessel aging is of course linked to irradiation damages: The main influencing factors are the degree of irradiation of the vessel and the loadings sustained during power transients. The operator minimises the irradiation* of the vessel by using fuel loading plans optimised from the neutronic point of view. Knowledge of the condition of the vessel material, in particular on the internal side is essential because existing defects may, depending on their size, favour the propagation of cracks. Experimentally, the irradiation of the vessel is monitored by means of dose measurements on test specimens. Also, the Nuclear energy of the future: what research for which objectives? 51

50 vessel s condition is checked by ultrasounds, which enable the size of the defects linked to cold fissuring and those resulting from intergranular decohesions caused by heating to be detected and evaluated. The CEA is deeply involved in a large R&D programme which accompanies this programme on the vessel s lifetime. It covers the main influencing factors regarding the evaluation of the vessel s strength and its lifetime. The main R&D programme concerns the Fig. 55. Vessel defect inspection machine. physical criteria justifying the strength of vessels. Apart from the evaluation of the fluence* sustained by the vessel, the programme comprizes amongst other things: Irradiations of steel test specimens in experimental reactor vessels (Osiris); The development of methods for the determination of mechanical properties; The development of advanced methods in fracture mechanics (probabilistic methods), aiming to better evaluate available margins of resistance. The monitoring of the condition of the vessel material via non-destructive testing methods is the subject of R&D actions, in particular regarding the improvement and the qualification of the ultrasonic processes implemented. The operating return from reactors shows a few aging phenomena that must be taken into account in order to be able to assure the lifetime of the containment. It constitutes the last barrier for the retention of radioactive materials in the event of a serious accident. In order to justify an increase in the lifetime of the containment, it is necessary to show that it would still play its role in an accidental situation. The various aging phenomena observed or envisaged are corrosion of the internal metallic skin and the degradation of the concrete containment, by fissuring or corrosion of the reinforcements. CEA contributes via its R&D programmes to the improvement of knowledge regarding these subjects. The aging of replaceable components CEA does not carry out specific programmes on the aging of all of the replaceable components from reactors. However, given their importance, some components are the subject of special attention. This is the case of steam generator tubes, the rupture of which may have serious consequences. The R&D programmes carried out at CEA concern non-destructive testing methods applicable to these tubes, and the two main aging mechanisms identified: corrosion under stress and wear and tear by friction due to flow-induced vibrations. The vessel internal components are also the subject of special attention, with the study of the hardening of steels due to irradiation, and the corrosion under stress accelerated by irradiation. R&D programmes on the subject result, in particular, in the irradiation of internal materials in fast neutron reactors. The wear and tear of control clusters, cluster guides and control mechanisms has been noted on the fleet, and is also closely monitored. The mechanism identified is tribo-corrosion, which associates wear and tear and passivation depassivation cycle of the oxidized metal surfaces. This programme associating the physical chemistry and the mechanics must lead to the understanding of these phenomena, their modelling and the production of rules to evaluate the aging and the control cluster replacement policy. In conclusion, it is important to specify that the calculation methods in solid mechanics, in particular in the field of fracture mechanics, have made such progress following the computer and digital revolution (finite element analysis) that we are currently better equipped to predict the detailed behaviour of the installation without having to resort to unfavourable simplistic hypotheses. If, at the present time, we find ourselves able to foresee a longer component lifetime, it is largely due to the modern techniques of digital computing. 52 The near future: research supporting the existing nuclear power

51 Preparing the replacement of current reactors with more efficient and safer 3 rd generation reactors Firstly let s briefly recall the various generations of reactors since the 50s: The first generation of reactors The first generation of reactors was strongly influenced by fuel cycle constraints, particularly in the 50s and 60s, with the absence of uranium enrichment industrial technology, and on the other hand with the willingness of some nations to equip themselves with a nuclear deterrence tool requiring the production of fissile materials. In this context, reactors had to be able to operate with natural uranium (non-enriched, requiring the use of moderators such as graphite or heavy water*.this is why family of Natural Uranium Gas-Graphite reactors (GGR), was developed in France. Three reactors, intended for producing plutonium (G1, G2 and G3) were created first, then six others intended for generating electricity (Saint-Laurent 14, Bugey 15 and Chinon 16 ). CEA was very strongly involved in the development of this system, in the capacity of process supplier. The Magnox type reactors in Great Britain belong to the same generation.these reactors presented interesting characteristics (thermodynamic efficiency, optimised use of uranium in the reactor core, etc.), but also limitations linked to the technology of these types of reactors, in view of development on a much larger scale: high investment cost, difficulty in improving the safety and the extrapolation to a much higher capacity, which penalized their economic performances as compared to light water reactors. This first phase saw a rise in concerns developed relating to the fuel cycle, regarding both the rational and sustainable use of natural resources (recycling of energy materials, in particular plutonium) and the question of waste management. This led to the development of processes and installations for the back-end of the fuel cycle: reprocessing of spent fuel, recycling of plutonium. From the beginning, France thus adopted the fuel cycle based on reprocessing-recycling, enabling on the one hand a better use of resources, by recycling plutonium in the reactors, and on the other hand, the reduction in the quantity and the long-term harmfulness of final waste, conditioned in order to ensure safe and sustainable containment of the radionuclides. The first UP1 reprocessing plant in Marcoule, for the reprocessing of GGR fuels, was commissioned in 1958, followed by the UP1 plant in La Hague in 1966, itself equipped in 1976 with a new workshop (HAO) for the reprocessing of pressurized water reactor fuel. They are now replaced by the two UP3 (1989) and UP2-800 (1994) plants in La Hague. MOX fuel manufacturing installations have likewise been developed and commissioned: CFCa Cadarache ( ), Dessel in Belgium (MOX fuels produced from 1986) and Melox in Marcoule (1995). The second generation of reactors The second generation of reactors which corresponds to the majority of the global fleet currently in operation originated from the need to render nuclear energy more competitive and from the willingness to reduce the level of energy dependency of certain countries at a time where a great deal of tension on the fossil energy market was being felt. The production of fissile materials for defence purposes was no longer a priority, enriched uranium produced by gaseous diffusion was commercially available (Eurodif plant in France). This period was that of the deployment of water reactors, pressurized water reactors PWR and boiling water reactors BWR, which constitute more than 85% of the current global electro-nuclear fleet of approximately 450 reactors. Industrial feedback from the last few decades has enabled economical as well as environmental performances of the production of nuclear energy to be demonstrated, with a highly competitive cost of the nuclear kwh in relation to that of fossil energies and a continuous reduction of waste and effluents, well below the authorized limits. The cumulated operation of more than 10,000 reactor-years proves the industrial maturity of this technology. 14. Municipality of Saint-Laurent-Nouan (Loir-et-Cher). Nuclear power station on the Loire (Note of the Editor). 15. Municipality of Saint-Vulbas (Ain). [Note of the Editor.] 16. Indre-et-Loire (Note of the Editor). Nuclear energy of the future: what research for which objectives? 53

52 The third generation The third generation represents the most advanced constructible industrial state-of-the-art. It involves reactors known as evolutionary : They benefit from the feedback and the industrial maturity of second generation water reactors, whilst integrating the most advanced specifications in terms of safety, knowing that the second generation already shows a very high level of safety. 3 rd generation reactors are the subject of a large international offer. These reactors are already being constructed in particular in Asia, but also in Finland and soon in France The types of 3 rd generation reactors Advanced water pressurized reactors AP 600, AP 1000, APR1400, APWR+, EPR Advanced boiling water reactors ABWR II, ESBWR, HC-BWR, SWR-1000 Advanced heavy water reactors ACR-700 (Advanced CANDU Reactor 700) Small and medium power integrated reactors CAREM, IMR, IRIS, SMART Modular high temperature gas reactors GT-MHR, PBMR Renewal at 50,000 MW spread over 30 years ( ) Rate of nuclear construction: 1,667 MW/year Installed power (MWe) Current fleet 40 years lifetime Extension beyond 40 years Generation Generation Average fleet lifetime: 48 years Fig. 56. The renewal schedule of the French nuclear reactor fleet, as currently envisaged by EDF. The operator will without a doubt wish to extend the lifetime of the existing reactors as much as reasonably and legally possible. It is planned to start the replacement of one part of the fleet in bevel as of 2020 in order to smooth out the financial effort, firstly with third, then fourth generation reactors. France is extensively equipped with nuclear power and its fleet of reactors is relatively young.yet, the construction of a demonstration EPR has just been decided. Why is this step being taken now? The development of a new system is a long-drawn-out operation: in order to introduce third generation reactors in 2020, it is necessary to order an EPR prototype now. The schedule envisaged for the EPR deployment in France is as follows: 2005 Decision for an EPR demonstrator Regulatory authorisation process and preparation for construction Construction and commissioning of the EPR demonstrator Acquisition of the operating feedback (minimum 3 years) 2015 Decision to construct an EPR series (number and rate to be defined) 2020 Commissioning of the first reactor of the series Commissioning of the following reactors 54 Preparing the replacement of current reactors with more efficient and safer 3 rd generation reactors

53 CEA is associated with the EPR*, pressurized water reactor, prototype studies which represent the fruit of approximately 10 years of collaboration between Framatome and Siemens. The two industrialists designed a reactor which uses the best of the technologies from French N4 and German KONVOI reactors. Industrial issues The EPR meets two main objectives: To make nuclear energy more competitive in relation to fossil energies; To further strengthen the reactor safety. A technical lifetime of 60 years, compared to 40 years in general for current power plants. The reactor should be able to operate for 40 years without important rejuvenating operations; Reduced operating costs: increased availability approaching 92% compared to 82% today, partly due to shorter shutdowns for reloading (in the order of 16 days) and to design choices (simplified maintenance of components is made possible during operation with the aid of the redundancy of safety circuits), reduction of the collective radiation doses for the maintenance staff (0.5 compared to 1 m.sv/year currently); An optimised construction time (approximately 57 months); Containment designed to resist a hydrogen explosion Molten core (corium) retention device in the event of an accident Heat evacuation system Strengthened safety combined with a more forgiving system regarding possible control faults, a significantly improved in-depth defence regarding the resistance to possible serious accidents (core fusion). The benefits provided by this strengthened safety results in the non-necessity of evacuating populations, even in the event of a serious accident. Water tank Fig. 57. The EPR project: a pressurized water reactor design which takes into account a large amount of feedback from second generation PWRs, with increased safety requirements. Main characteristics of the EPR The characteristics of the EPR, enacted by an omnipresent concern to improve performances and economy, may be summarized as follows: A net electrical capacity of approximately 1,600 MWe (compared with 1,450 MWe of the N4), well-suited for regions with many well-linked power grids.this increase in capacity downscales the costs per KWh; An energy efficiency of approximately 36% (10% better than reactors from the previous generation) mainly due to increase in performances of steam generators and turbines; A possible use of various types of fuel (UOX* or MOX, or even 100% MOX) thus allowing a flexible and more economical management of resources and waste (15% reduction of the quantity of uranium needed to produce the same quantity of electricity); Nuclear energy of the future: what research for which objectives? 4 independent areas for the redundant safety systems R&D objectives and challenges The EPR enables optimised reactor operation management and a higher degree of flexibility of use of the fuel resulting directly in a better competitiveness. In terms of safety, an important effort has been carried out in order to minimize the consequences of possible core melting accidents thus contributing to better public acceptance. These two topics relating to the fuel and the safety are still important R&D challenges in order to improve the concept and make it yet more competitive. In the near future, CEA will mainly intervene in the field of physics and core management and in the field of safety. Core physics and EPR core management EPR reactor cores are made of the same standard fuel elements as pressurized water reactors.they mainly differ in size, a slightly lower fuel rating and a more economical fuel management in fissile material. 55

54 CP1 CP2 P4 P 4 N4 EPR Electric power 900 MW 1,300 MW 1,450 MW 1,600 MW Core thermal power 2,785 MW 3,817 MW 4,250 MW 4,450 MW Type of assembly 17*17 17*17 17*17 17*17 Number of fuel rods per assembly Fuel rod height (m) Number of assemblies Average linear power density (kw/m) A heavy reflector, made of a steel plate surrounding the core, enables a better neutonic economy. Thus the core is more reactive and will require a lesser investment in fissile material in order to produce the same quantity of energy.this gain also shows in cycle length. The heavy reflector also enables a reduction of the high energy neutron flux* on the vessel. This reduction of fluence authorizes as of today a vessel lifetime of 60 years. An experimental as well as theoretical R&D programme is required to increase the accuracy of the industrial calculation schemas and save the maximum amount of fuel. In parallel, it is also necessary to further qualify the neutron transport calculations of the fluence on the vessel. The large size of the EPR s core requires three-dimensional calculation methods with local reconstruction of the power. They require a specific qualification from the neutronic point of view (for the accurate evaluation of strong gradients of neutron flux in the fuel bundles), and from the thermal-hydraulic point of view (in order to correctly evaluate the local development of the moderation by the water which causes the neutron transient. EPR fuel management modes UOX managements For UOX cores, the standard management mode projects a stay time of the fuel in the reactor of 72 months, with renewal of a quarter of the fuel every 18 months, the final burnup rate being 60 GWd/t. New 1 cycle 2 cycles 3 cycles MOX management modes Heavy reflector Water Core envelope Water Fig. 58. Positioning the UOX fuel bundles in the EPR core. The benchmark management modes for the use of MOX fuel are 50% MOX hybrid modes with renewal by a quarter of the UOX bundles and by a third of the MOX bundles. The length of the cycles is 18 months, the respective burnup rate reached by these two types of fuels are 60 and 55 GWd/t. The fuel management modes in the reactor are chosen to save fuel, by guaranteeing reactor shutdown times as short as possible, by assuring a long lifetime to the various reactor components, all this in compliance with safety rules. These management modes therefore result in a complex optimisation, the result of which narrowly depends on the characteristics and performances of the fuel and the components of the core itself. The continuous improvement of these performances already leads to innovative management modes being researched for the EPR reactor. Apart from the conventional studies carried out and necessary for the neutronic qualification of the desired burnup rate calculation schemas, which consist of: Analysing the isotopic composition of the irradiated rods; Evaluating by oscillation in the Minerve reactor the integral cross sections in order to evaluate the uncertainty of the calculation schemas and to minimise it; Optimising and validating the neutronic calculation schemas for these burnup rates; The specific design of the EPR core requires a certain number of additional qualification elements. This core can accomodate various reflector concepts and exact evaluations of their influence are carried out there as much on the core s neutronic properties, as on the fission distributions in the core s rods, in particular in the periphery. In addition, this programme must enable the measurement of the neutron flux and in particular fast neutrons beyond the core in order to be able to validate the calculation of the vessel s fluence. 56 Preparing the replacement of current reactors with more efficient and safer 3 rd generation reactors

55 The fuel without PCI A main objective is to eliminate the constraints due to the class 2 PCI (pellet-clad interaction) by Candidate products exist (in particular the UOX fuel doped with chromium). Three study points were retained: The implementation of irradiation tests in more demanding transient operating conditions (power transients simulating the load monitoring) for the qualification of the fuel; The 3D fine modelling aiming to obtain a better understanding of the PCI phenomenon. A relevant physical modelling must enable the evaluation of the damage leading to the cladding s loss of integrity, the impact of the design s developments, and the impact of the operating and transient history; To study the accidents which must inevitably be taken into account with the new core managements and new fuels (for example, rupture of steam piping system) thermal-hydraulicneutronic-fuel coupled calculations are necessary.these studies are carried out with the current calculation codes, Cathare and Flica for the thermal-hydraulics, Meteor for the fuel and Cronos for the neutronics. For the future (by 2010), these analyses will be conducted more easily with the calculation tools in process of co-development by CEA and its industrial partners: Neptune for the thermal-hydraulics, Descartes for the neutronics and Pléiades for the fuel, tools integrated into a unique software platform. The 1.5D industrial modelling copied both on the tests and on the above calculations. The EPR s safety The evaluation of the consequences of a serious accident on the EPR is placed within the framework of the in-depth defence safety procedure and the joint recommendations issued by the French and German Safety Authorities published in The procedure established as of the design studies according to a deterministic method complemented by probabilistic studies targeted the following objectives: Elimination in practice of the accidental conditions which may lead to important discharges of fission products in the shortterm; Elimination of the need to displace the population in serious accident situations, without the emergency evacuation of the close neighbourhood and without long-term restriction for the consumption of food products. In serious accident situations on the EPR, the abovementioned objectives may be obtained using a strategy enabling the integrity of the containment to be assured. The strategy mainly relies on the possibility of reliably depressurizing the primary circuit, on the establishment of hydrogen recombiners in the containment, on the installation of a double-wall containment with filtration in order to reduce the risks of radiation leakage and finally on the design of a corium catcher responsible for assuring stabilisation of the corium over the long term. Fig. 59. Over the past approximately twenty years, CEA, EDF, IRSN 17 and Framatome-ANP 18 have developed the Cathare accidental thermal-hydraulics code. This tool enables any type of accident that may occur on light water reactors to be simulated. More particularly, the Cathare code presents a wide validation field for PWR accidents; it is used intensively in France by industrialists and the Safety Authority for all of the case studies relating to the safety and control of reactors. With reference to the development of the EPR, the code was used as a benchmark tool for the design and for accident studies. For the EPR, primary coolant loss accidents of the large break type (APRP-GB) are in principle excluded by design; indeed, leak detection devices on the primary piping enable them to be avoided. In serious accident situations with important degradation of the reactor s core, the mixture of molten materials (corium) would eventually attack the bottom of the vessels and could pierce the wall. In order to collect and stabilize the corium over the long-term, a spreading type of retention device has been 17. Institute for nuclear radioprotection and safety (Note of the Editor). 18. Framatome ANP stems from merging of the nuclear activities of Framatome and Siemens. Framatome ANP has built more than 90 reactors, more than one third of the world s nuclear capacity (Note of the Editor). Nuclear energy of the future: what research for which objectives? 57

56 planned underneath the vessel (Figure opposite). This corium catcher system is an innovation with respect to current reactors. It should be noted that numerous studies in the fields of high temperature metallurgy, physical chemistry of materials and rheology* were conducted in France and Germany in order to design and test the EPR s retention device. The EPR s corium catcher Sacrificial material Protective layer: Spreading compartment Sacrificial material -7.80m Basemat cooling Zirconia layer Melt discharge channel Melt plug Fig. 60. Framatome ANP has designed a corium catcher outside of the vessel based on a concept of spreading over a large surface area with long-term cooling and stabilization of the corium. The retention device is located in a dedicated compartment in the containment so as not to sustain important stress during the vessel s rupture. This compartment is separated from the reactor pit by a melt door. In order to deal with the long-term situation, it is necessary to be able to evacuate the residual power (in the order of 35 MW) for a corium mass of approximately 200 tonnes. The cooling efficiency - in the upper part thanks to the corium flooding, and in the lower part thanks to the metal cooling structure - will enable all of the corium to be stabilized within a few hours and solidified within a few days. 58 Preparing the replacement of current reactors with more efficient and safer 3 rd generation reactors

57 Research regarding waste management Whereas large volumes of short lived radioactive waste are currently managed industrially in surface disposal sites, the long-term management of long-lived high level radioactive waste is the subject of important research in the countries generating nuclear electricity in significant quantities, such as France, Japan and the United States. What should be done with this waste has remained a conflicting question for many years. Yet, scientific knowledge is advancing and technical solutions are taking shape. Nevertheless, science and technology largely interfere with the social dimensions of the problem posed. People s fears remain strong and difficult to appease, especially when the danger persists on time scales which defy common comprehension. In addition, the geopolitical context, the energy crisis which has followed and the fears of consequences of climate warming open ideological debates regarding the energy choices to be made and the very nature of the economic development to be privileged for a sustainable exploitation of the planet s resources. Nuclear energy and radioactive waste management play a large part in these debates. What is the future for this long lived radioactive waste? In France, the Bataille Act, voted in 1991, clearly posed this question to the scientific community by requiring that all options be examined, and by proposing several research avenues. What are the results of these efforts and what prospects do they offer? The orientations of this research are mainly concerned with reducing the volume and dangerousness of the waste by sorting and recycling. These are the same principles as those retained for the management of other household and industrial waste. They have been implemented for several decades with the industrial reprocessing of spent fuel in La Hague, which enables energy materials that can be upgraded, such as plutonium and uranium to be recycled. Can we do better? This is the question posed to scientists. The radioactivity of waste, which for some may persist for long periods, requires the use of effective containment systems as long as the danger subsists. The cost of these diverse measures, evaluated by the yardstick of their effectiveness, will of course have a decisive impact with regard to decisions and to the schedule which will be implemented resulting from the law. Nuclear energy of the future: what research for which objectives? Several lessons are currently confirmed The very nature of final radioactive waste, which cannot be recycled or reused, depends on the available technology: the final waste in 30 years may differ from that produced today. Thus B and C (glass) waste, already produced, is final waste for our generation. Eventually, it will without a doubt be possible to further reduce the radiotoxicity* of vitrified C waste by eliminating some of the radionuclides (minor actinides) that it still contains today. This is the subject of the research on enhanced partitioning and transmutation. The elimination of these radionuclides would also reduce their thermal power. Nevertheless it will be necessary to be able to make them disappear by transmutation if a net gain in the radiotoxic inventory is to be produced. Prior to becoming an industrial practice, these technologies require complementary research and development in order to enable their integration into a viable economic contex; A permanent host location must also be found for the final waste. Deep geological disposal seems to be the only very long-term management solution where the safety measures do not require continuous control by the society. An international consensus was established on this question, agreed on by the International Atomic Energy Agency (IAEA), the Nuclear Energy Agency (NEA) or the Organisation for Economic Co-operation and Development (OECD). No other equivalent solution has appeared, neither in France, nor elsewhere in the world; A geological disposal facility will always be a rare, therefore expensive, resource. It should be used as effectively as possible, by further reducing the volume and the thermal power of the final waste which will be stored, two parameters which largely condition its capacity, therefore its utilisation period, and its cost. The reprocessing of spent fuel, practiced in France, is already going in this direction because it enables the extraction of uranium representing more than 90% of its mass, and the recycling of plutonium, the highest contributor to its overall radiotoxicity; Therefore, the American AFCI (Advanced Fuel Cycle Initiative) initiative is exemplary. After twenty years of efforts leading to the decision of creating the spent fuel disposal site of Yucca Mountain, the Americans are now considering the optimization of its usage, and therefore the very nature of the objects which will be placed there; 59

58 In addition, waste conditioning studies will be continued by researching an even better containment and a volume reduced as much as possible. The space thus saved in the disposal site will help to make the industrial practice of increased spent fuel reprocessing profitable; uncertainties associated with long timescales.this option will only be practicable when we have the ability to separate and transmute radionuclides. Enhanced partitioning: what are the consequences regarding long-term waste management? The issue of enhanced partitioning, that is the additional extraction of radionuclides other than plutonium and uranium, is the reduction of the radiotoxicity of the future high-level waste. Figures 63 and 64 illustrate the decrease in radiotoxicity of UOX spent fuel as a function of time, as well as the relative contribution of each category of radionuclide to the overall radiotoxicity. Fig. 61. The underground laboratory operated in Bure by the French National Radioactive Waste Management Agency (ANDRA) Sv / tonne 6 Crest North portal South portal 5 4 U Other Main tunnel (ESF) 3 Pu Storage gallery 2 Emplacement Drift 1 PF Time in years Waste package Drip shield Fig. 63. Radiotoxicity of spent fuel Sv / tonne Fig. 62. The concept of the underground disposal site at Yucca Mountain (Nevada), developed by the United States Department of Energy. 1 0,9 0,8 0,7 0,6 0,5 Other Pu U Am Np 0,4 0,3 The safety of the disposal site relies on its capacity to contain radionuclides in geological formation until their radioactivity has sufficiently decayed. Finally, the demonstration of the safety of a disposal site will rely on the firm conviction of the correct operation of the installation. Studies must be pursued in order to better understand the evolution of the waste packages in disposal situations over time and the migration of radionuclides into the geosphere. Eventually, removing from the vitrified waste packages the long lived radionuclides which contribute the most to their long-term radioactivity, will significantly reduce the time during which they remain dangerous and will also have the effect of reducing the scientific 0,2 0,1 PF Time in years Fig. 64. Distribution of contributions to the spent fuel radiotoxicity. 60 Research regarding waste management

59 Plutonium* contributes approximately 50% of the initial radiotoxicity and 90% one hundred years later. Thus, as soon as the spent fuel has been processed, that is, the plutonium (with uranium) that it contains has been extracted, the residual radiotoxicity remains dominated by that of the fission products (FPs) and of curium over approximately one hundred years and, over a longer time frame, by that of other minor actinides (MAs) (americium and neptunium). The FPs and MAs are currently all incorporated into vitrified waste resulting from the reprocessing of spent fuel. A following stage would therefore consist of only including FPs in future C waste. The figure 65 shows the comparison of the decrease in radiotoxicity of the materials respectively contained in a spent fuel assembly, in a vitrified package produced today (MAs and FPs) and in a vitrified package from which the minor actinides would have been eliminated (thus only containing FPs). Relative radioactivity 10,000 1, Glass without MAs (FPs only) Conventional glass (MAs + FPs) Also the final purpose of the partitioning should be specified, that is the method of eliminating partitioned radionuclides. Which radioelements should be partitioned and which partitioning modalities should be retained: should radioelements be partitioned separately or as a group? Which purity and which chemical form must be given to the partitioned elements in order to meet the constraints of the following stages leading to their recycling and/or their permanent elimination? What consequences can be expected on the design of the geological repository site, its cost and its long-term containment performances? The answers to these questions are of course linked to the future of partitioned elements. The last stages of development may include an industrial process pre-control phase. Apart from minor actinides, partitioning studies have concerned a selection of FPs (iodine, technetium and caesium). Some FPs, which are much less radiotoxic than MAs, show particular mobility in the geosphere. Their lifetime, very long for some, such as isotope 129 of iodine, may pose the threat of a return to the biosphere in the very long term. However, the potential damage of these low radioactive radionuclides is very low, which currently authorizes the discharging of iodine into the sea. Spent fuel (Pu + MAs + FPs) 100 1,000 10, , ,000 Fig. 65. Decrease in the relative radiotoxicity as a function of time. (The radiotoxicity of the glass or spent fuel is estimated here in relation to that of the uranium which produced it.) Enhanced partitioning studies 19 currently concern the minor actinides americium, curium and neptunium. For the hydrometallurgical method, they are at the technical demonstration stage and already deal with a significant quantity of spent fuel (approximately 15 kg placed in solution). The process steps retained for the enhanced partitioning are largely based on industrial knowledge regarding the Purex process used in La Hague. A first technical and economical evaluation of an enhanced partitioning workshop will provide clarification on the technical modalities and economical conditions of use of such a process. ALARA* type criteria should guide future decisions regarding the matter. Enhanced partitioning must therefore mainly concern minor actinides in order to reduce the radiotoxic source term. Time (years) Beyond aqueous method processes, studies of processes based on pyrochemical techniques must be comprehensively pursued. The latter effectively present a potential advantage in terms of compactness. Partitioning is obtained in one pass and may concern highly radioactive spent fuel. These characteristics make up a technique which, if these advantages were confirmed in regard to the disadvantages (corrosion, partitioning efficiency, etc.) and if it reached an industrial development stage, could take its place in an integrated cycle on the reactor site, thus avoiding the transportation of radioactive materials over long distances. However, it is advisable to pay close attention to the secondary waste, salts and technological waste, which may result. The combination of pyrochemical and hydrometallurgical processes may also be envisaged. This technical demonstration alone will, however, not be sufficient to enable the industrial application of the process. 19. See infra, p. 67, the chapter entitled The fuel cycle of future nuclear systems. Nuclear energy of the future: what research for which objectives? 61

60 Processes for the partitioning of radionuclides Plutonium recycling is already an important stage for reducing the radiotoxicity of spent fuel. Beyond that, removing minor actinides from vitrified waste produced today would enable their radiotoxicity and their thermal power to be further reduced. The recovery of minor actinides may be a step toward their elimination by transmutation. The consequences on all of the steps of the fuel cycle must be carefully evaluated. Hydrometallurgical processes may be preferred, because they are better known. However, the potential of pyrochemical processes must continue to be examined. Nevertheless the question remains of knowing at which moment to begin the enhanced partitioning method, in coordination with the opening of reactor transmutation possibilities. Finally the net radiotoxic inventory which would result from implementing enhanced partitioning must be established. The economy of the process must also be specified in an ALARA context. The potential of various reactor systems for transmutation has been studied. It emerges that only the fast spectrum reactors, or systems combining a sub-critical* core reactor with an accelerator (ADS, Accelerator Driven System* 20 ) would enable transmutation efficiency to be obtained creating a real difference in radiotoxic inventory. Nevertheless, transmutation remains a very complex and certainly expensive operation, which cannot be applied to all radionuclides. Major technological questions are posed. Their resolution may only be envisaged in the context of the sustainable development of nuclear energy and the development of new nuclear energy production systems of the fourth generation. These questions concern all of the stages of future cycles: Which fuel and which waste reprocessing and recycling processes to use? What types of reactors? To date, it is admitted that a fast neutron reactor is the most performant tool for obtaining the transmutations of the elements envisaged. What is the economic impact for the entire cycle? The transmutation of partitioned elements: future cycles Transmutation is the operation by which highly radiotoxic radioelements are transformed into other elements with reduced or zero radiotoxicity. The research carried out for more than 10 years provides numerous clarifications and in effect confirms the possibility of reducing the radiotoxic inventory present in spent fuel. But this can only take place at the cost of important technological and financial efforts, concerning all of the nuclear fuel cycle plants and reactors, if a significant net gain in the radiotoxic inventory is to be recorded. How useful would it be to transmute MAs if the disposal facility reaches its currently calculated containment performances? Transmutation, in such a context, could appear as an additional measure of safety in the eventuality of a premature loss of the disposal site s containment. Transmutation, implemented in this way, will lead to final waste which it will be necessary to dispose of. All elements are not easy to transmute and the multiple recycling of plutonium and minor actinides, according to the scenario envisaged, could lead to the production of increased quantities in minor actinides, of which in particular, curium. This highly radioactive and very hot element poses management problems which remain yet to be solved. The Generation IV fast reactor route would enable the generation of electric energy and a better use of the natural resource uranium in the future by converting uranium 238 into a fissile element, via neutron capture. An ADS, dedicated to transmutation and generating no electricity, would be a more expensive and even more complex machine because it involves the coupling and the stable and reliable operation of two sophisticated components: a high intensity accelerator and a sub-critical reactor core. The time scale for implementing these new systems could be between approximately 30 and 40 years according to the means that are attributed to them. The continuation of transmutation studies will be carried out using experimental irradiations of various materials in fast reactor cores (Phénix* reactor), and on progressively increasing quantities. It is on the international level that they may progress the best by sharing experimental means, in France, Japan and Russia. The studies of scenarios will enable the best possible technological combinations to be identified in view of optimizing material inventories. 20. See infra, p. 95, the chapter entitled: Other avenues for the distant future: thorium cycle, hybrid systems, fusion. 62 Research regarding waste management

61 acterised by the emission of a thermal power in the order of two kilowatts on the date of their production; Standard compacted waste packages (SCWP) contain long lived intermediate level waste. It mainly concerns metallic elements of the structure of spent fuel bundles (tubes, braces, grids); Long lived intermediate level technological waste packages, cemented because not compactable, representing only 0.1% of the initial activity. Fig. 66. The Phénix fast reactor on the Marcoule site (Gard). The transmutation of minor actinides The transmutation of minor actinides could enable the reduction of the radiotoxicity of radioactive waste produced as of the advent of 4 th generation fast reactors, around 2030 to Prior to their implementation, many technological deadlocks must be removed regarding their design, their fuel (type and manufacturing and reprocessing technologies), their nuclear materials inventory (recycling of radiotoxic materials), their safety and their economy. Fast reactors would enable the use of large stocks of depleted and/or reprocessed uranium, thus conserving the natural resource. There are even more technological deadlocks to be removed for the ADS. The conditioning of waste and the long-term behaviour of the packages The conditioning of B and C radioactive waste, originating from the reprocessing of spent fuel, has been carried out continuously according to the industrial standards approved by the Safety Authority, since the commissioning of the UP3 facility in La Hague. It gives the waste a physical and chemical stability which prevents its dispersion into the environment. At the end of the chain, it leads to the production of packages enabling the waste to be easily handled. Fig. 67. Standard vitrified waste container (SVWC). Fig. 68. Standard compacted waste container (SCWC). Three types of waste packages are currently standardized: Standard vitrified waste packages (SVWP) contain the quasitotality of the initial radioactivity of spent fuel. They are char- Nuclear energy of the future: what research for which objectives? Fig. 69. Bitumined waste package. Fig. 70. Technological waste package. 63

62 The oldest C waste, originating from the industrial cycle or from research, was, either stored whilst awaiting its vitrification, or already packaged in glass; B waste was transformed in forms different from current standards, in cement or bituminous packaging matrices. For most of this old waste, producers have already specified the benchmark strategies that they intend to follow in order to condition them. R&D actions are carried out on the benchmark conditioning. This is for example the case of STE3 sludges in La Hague, for which COGEMA plans to use a bituminous packaging, suitable for the specificities of these sludges. For existing packages industrially produced according to current standards, the research aims above all to evaluate their durability, or more generally their long-term behaviour, in storage and then disposal conditions. The results of these studies thus contribute to the evaluation of the long-term containment performances of radioactive waste management modes, and therefore the safety of the latter. The quality of the conditioning carried out contributes to delaying the moment from which radionuclides start to migrate out of the package. Behaviour studies are thus interested in all of the packages and most particularly in vitrified waste packages. The durability of vitrified waste packages has been studied for over twenty years. It will continue to retain attention because this package contains the highest radioactive inventory. Glass is currently the most durable material (matrix) used industrially in order to host and immobilize a large inventory of highly radioactive radionuclides. Its behaviour over a few hundreds of years in storage conditions poses no problems. Over a much longer term and in geological disposal conditions, studies have already enabled the mechanisms and kinetics of alteration in question to be identified. Studies still in progress aim to better determine the moment from which radionuclides may start to disperse outside of the package into the rock of the geological formation, then into the geosphere, to then reach the biosphere. While the French strategy consists of reprocessing spent fuel, studies on the long-term behaviour of spent fuel have nevertheless been carried out in the hypothesis of long-term storage or even geological disposal. According to the decisions which will be taken at the end of the 15 year-period prescribed by the French law, these studies could be stopped or redirected to very specific aspects. The research on the conditioning and long-term behaviour of waste packages must be continued in order to accompany technological developments, as for example the increase in the burnup rate of spent fuel envisaged by EDF, which will result in modifications of the type and quantity of radionuclides which are found in vitrified waste. Finally, new conditioning matrices have been studied in order to contain over long periods elements that are difficult to transmute and/or elements that are particularly mobile in geological disposal conditions. It will be advisable to evaluate the relevance of their implementation regarding the particular risk that they may help to reduce and the overall cost that this implementation could represent. Nothing to date lets presume that such an option must in principle be retained. Grain boundaries +Mo, Tc, Ru, Rh, Pd 14 C 129 I Fractures Cladding-pellet gap Inter-pellet gap Gap 135 Cs 137 Cs 79 Se 99 Tc 90 Sr 36 Cl Closed porosity Matrix Actinides Fission products (98 %) Zr cladding 14 C 93 Zr 36 Cl Fig. 71. Glass casting in the laboratory at Marcoule (Gard). Fig. 72. Simplified diagram of the microstructure of spent fuel and the location of the various radionuclides. 64 Research regarding waste management

63 The containment of the radioactive materials The conditioning of waste consists in the production of packages which assure the containment of the radioactive materials and makes their handling possible. The research on conditioning and the long-term behaviour of waste packages remains open, in order to accompany the technological developments in progress (increase in burnup rates, etc.) or the decisions expected by the Bataille act deadline in The relevance of using conditioning specific to a given element must be evaluated. Studies on the long-term behaviour of packages, in particular of glass packages in disposal situation, must be continued in order to confirm the demonstration of safety of geological disposal. accessing information at any moment regarding the packages which are stored there. In future, the studies will concern rather the way of best mastering the durability of the concrete and materials used in the storage installations. If in 2006 the decision was made to engage means in view of the creation of a geological repository site, it would then be advisable to optimise the thermal management of the thermogeneous packages, which put a strain on the cost of such an installation. It will be necessary to accurately evaluate consequences of a prolonged storage of these packages, enabling their cooling in economical conditions, prior to their permanent disposal and to then determine the temperature acceptable by the host rock, once the site is known. Finally, according to the actinide partitioning modalities which could be envisaged, it would be advisable to continue the studies on the storage of grouped or partitioned radioactive materials, for example in the case of curium, in order to assess its feasibility and to evaluate its consequences on the fuel cycle. -40 m Model pit for spent fuel storage Room 2 Room 3 8m 6m 31 m 8,3 m Fig. 73. Upper part of the storage wells of the CASCAD installation in Cadarache (Bouches-du-Rhône). Storage: a pending solution The most radiotoxic types of radioactive waste, B and C, are all stored to date in industrial installations operated by waste producers pending a final destination. Fig. 74. Galatée: a demonstration gallery of sub-surface storage suitable for the long-term, recently constructed on the CEA s Marcoule site. This 40 metre long structure, with 8 m by 8 m cross-section, shows the components of such a warehouse and the logistics of container handling. It will be used for thermal experiments in view of the validation of the behaviour models and codes of a concrete structure subjected to operating hazards, such as the loss of cooling. These installations operate without particular difficulties and their projected lifetime is approximately fifty years, with possible extensions beyond this limit. The studies on long-term storage, carried out within the framework of the 1991 act, have identified factors limiting the lifetime of such installations, in particular the alteration of concrete and the corrosion of metals. The risk of neglect by the society remains the intrinsic weakness of such an installation the safety of which, medium and long-term, relies on continuous monitoring and maintenance, and on the possibility of Fig. 75. The Galatée facility viewed from outside. Nuclear energy of the future: what research for which objectives? 65

64 The storage Storage is the management mode which makes it possible to await the arranging of an outlet for the final waste. The durability of the concrete and materials used in the warehouses, which limits their lifetime, may be the subject of complementary studies. The heat released by some packages puts a strain on the cost of the disposal site. The optimisation of the thermal management of these packages, is therefore to be undertaken as soon as the disposal site is known. The issues and the feasibility of the storage of materials originating from an enhanced reprocessing of spent fuel must be evaluated. Fig. 76. The Galatée facility viewed from inside. 66 Research regarding waste management

65 The fuel cycle of future nuclear systems The design of the processes involved in the fuel cycle relies on two main determining elements: firstly the choice of a matter management strategy (which must evidently be consistent with the capacity of the fleet of reactors to use this matter efficiently), and then the fuel object itself (its type, its composition, its morphology): fuel is the backbone of the cycle and its choice is closely linked to that of processes, involved in its fabrication and processing. Reflections carried out on 4 th generation systems are currently still very plentiful, as much for reactors as for their fuels, for which highly diverse configurations may be envisaged: oxides, carbides, nitrides, in the form of pins, particles, filaments, or even molten actinide salts. Such a profusion is at this stage absolutely normal, and even fortunate, but does not enable exact orientations to be made for the processes to be implemented; it will actually be the cycle studies associated with each of these concepts which may and must help decide on the choices in terms of fuel. On the other hand, concerning the matter management strategy, a few main orientation elements are already emerging from the reflections of the last few years on the international level, in particular within the framework of the Generation IV Forum. The main expectations and conclusions which seem to come out at this stage regarding the research directions to be privileged (also raising numerous questions, which currently remain open) are summarized below. Which material management strategy? The criteria which frame the reflection are those which are essential for nuclear systems of the future: sustainability, economic efficiency, safety, are the three main aspects by the yardstick of which the forum community has chosen to evaluate the various concepts that can be envisaged. Even though the fuel cycle must consider these three issues, the most important one is certainly the issue of sustainability, whether it concerns the preservation of natural resources, the minimisation of the environmental impact or the resistance regarding the risks of proliferation. or closed cycles with deployment of fast neutron reactors), the evolution of the quantities of residual heavy nuclei on the one hand (which is an indicator of resources called into the disposal site), and the natural uranium requirements on the other hand. It is established in an obvious way that the most enhanced recycling options are the most effective regarding the various components of the sustainability criterion: The upgrading of the energy potential of uranium 238 and the multiple-recycling of plutonium in fast neutron reactors is of course the key factor for the preservation of resources 21 ; The recycling of plutonium and minor actinides essentially contributes to minimizing the residual inventory in fissile nuclei, the potential noxiousness of waste and also its longterm thermogenous character. The first and main central idea for the cycle thus emerges from these reports: sustainable nuclear-power seems to require a recurrent and enhanced recycling of actinides. However, the Worldwide spent fuel Heavy metal mass (thousand tonnes) LWR once through LWR + fast reactor Year Fig. 77a et 77 b. Prospective elements. Water reactors rapidly consume fissile resources and accumulate actinides. Fast reactors do not present these defects. This becomes quite apparent when reading the graphs in Figure 77, presented during the forum s work. These graphs indicate for various scenarios (open cycle and water reactors, Nuclear energy of the future: what research for which objectives? 21. For an explanation of the capacity of fast reactor for efficiently consuming fertile materials such as uranium 238 and therefore for using the heavy metal resources, as well as possible, see infra p. 75, the chapter entitled: On the origin of species (of reactors): systems and generations. 67

66 Cumulative natural U (millions tonnes) Worldwide uranium resource utilization Speculative resources Known resources LWR once through Fig. 77 b. Prospective elements. The worlwide utilization of uranium resources depends on the type of reactors and fuel cycle, and on when they are put into use. options remain open regarding recycling scenarios, regarding the boundaries of all of the elements to be considered among the transplutonium elements (according to their inventory, their properties, their impact, the difficulties that their recycling may entail); but the general outline seems clearly traced and leads to the block diagram in Figure 78. Uranium Actinides Fast reactors introduced 2050 Fast reactors introduced 2030 Year PF The importance of retaining compact technologies, in order to reduce investment costs; this aspect becomes important if one choses decentralized processing options, with reprocessing and recycling on the same sites as the reactors. These options avoid the transportation of large quantities of spent fuel but lead to an increased number of plants with lower capacities; Finally, the necessity to favour implementation of clean technologies, that is minimising as much as is reasonably possible the effluents discharged and the (secondary) technological waste generated 22. Which recycling processes? Again in this field, the reflection may seem plentiful; as indicated above, the choice of a cycle process depends on that of a fuel, and it is obviously too soon to decide on precise options. But in order to orient or frame the reflections, a few general ideas mainly originating from industrial feedback or prospects offered by the advances in research may be brought up. The selective recovery of actinides by use of their physical properties does not seem currently to be taking off as anticipated in some respects. If indeed the objectives of a partitioning of the actinides from all of the fission products is considered, it in fact involves partitioning the heaviest nuclei from the others: this evidence has however not given rise up to now to the enhanced exploration of concepts based on field effects for spent fuel reprocessing processes. The aim of generalized recycling of all of the actinides may give fresh impetus to the research in this field, but the technological jump will most certainly be significant. Reactor(s) Processing Therfore, the current reflection is mainly centred around the potential of chemical processes, usually distributed between hydrometallurgical ( aqueous method) or pyrometallurgical ( dry method) processes. Fig. 78. Block diagram of the fast reactor cycle. The first have to their credit impressive industrial feedback: They make use of a mature technology. As shown by the results obtained with the PUREX process implemented in the This first point being defined, other orientations then emerge from the reflections carried out. Perhaps less obvious or less unanimous, they seem to say a lot about the questions posed: The importance of a grouped management of recycled actinides, which avoids the recovery of isolated fissile isotopes.this seems to make the recycling process more resistant to proliferation (by reducing both the strategic value of the materials for the applications concerned, and their accessibility thanks to the presence of highly radioactive nuclei). This also goes in the direction of improved economic efficiency, by simplifying the matter management processes; 22. Here we are mainly interested in the management of materials originating from a uranium fuel cycle. The hypothesis of deploying systems bringing thorium into play has been approached during expert meetings within the framework of the Generation IV forum: in spite of the potential interests of such systems in some respects (abundance of natural resources, lower generation of radiotoxic heavy nuclei), those do not seem to be put forward for the next generation of reactors (with the exception of the options studied of molten salt reactors, for which thermal spectrum breeding with thorium 232 can be envisaged); this essentially relies on the sustainability of the uranium resources in the hypothesis of an upgrading of uranium 238, on the privileged options of recycling all of the heavy nuclei, which reduce the question of the long-term radiotoxicity of the residues, and also on the impact of the accumulated experience regarding the uranium system. 68 The fuel cycle of future nuclear systems

67 La Hague plant, hydrometallurgical processes procure very high partitioning performances (recovery rate and purification factor of recycled materials) whilst resulting in a low flux of generated technological waste. In addition, they appear to offer a large potential to adapt (to fuel characteristics, but also to recycling specifications, as shown in the studies recently carried out on the complementary partitioning of minor actinides). They also display undeniable residual progress margins (in particular for increasing compactness, thus reducing the cost of their implementation). They therefore seem to be the benchmark method for the development of advanced cycle concepts for the fourth generation of reactors. Pyrometallurgical processes are currently presented as the main alternative to aqueous processes, and are the subject of a renewed development effort on the international level. The generic principle of such processes consists of placing elements to be partitioned in solution in a bath of molten salts (chlorides, fluorides, etc.) at high temperatures (in the order of several hundreds of degrees Celsius), and then of operating the partitioning of the species of interest via diverse techniques such as extraction via molten metals, electrolysis, or selective precipitation.the interest of this type of process mainly resides in the high solubilisation potential of ionic liquids (in order to dissolve refractory compounds), in the low radiosensitivity of the inorganic salts used (which would enable the on-line reprocessing of fuels to be envisaged as of their unloading), in their compactness (few successive transformation stages can lead to a recyclable product), as well as in the best aptitudes presumed for a joint management of the actinides. It is presented moreover as the unavoidable, natural process, of online reprocessing of liquid fuels from molten salt reactors. The Argonne 23 (see Figure 79) and Dimitrovgrad teams have carried out important developments on such concepts, respectively for the reprocessing of metallic or oxide fuels, up to the creation of pilot installations on which demonstration campaigns have been carried out. However at this stage strong uncertainties remain, the most noticeable concerning the level of partitioning performances (in particular the actinide recovery rate) and the implementation on an industrial scale of the technology (secondary waste generated, particularly given the aggressiveness of the operating environments and conditions). 23. Argonne National Laboratory. This American research body is supervized by the university of Chicago, for the department of the Energy of the USA (DOE). Nuclear energy of the future: what research for which objectives? Active metals + Spent fuel U Noble metals Cadmium Solid cathode Pu Cadmium liquid cathode Rare earths Molten salt Fig. 79. The Argonne pyrochemical an process consists of electrolysis in molten salt medium, with partitioning of the elements on the various components of the electrolyser. Which lines of action for research? Apart from highly diverse exploratory research, which may be carried out on concepts radically removed from the existing ones, a few large research avenues are taking shape relating to the two main previously mentioned concepts. Concerning the hydrometallurgical processes, efforts are orientated towards the following points: Firstly, adapting the current process to the characteristics of new fuels: this will mainly concern the dissolution of the fuel. The reagents and conventional dissolution conditions may prove to be unsuitable for certain advanced compounds. Earlier work carried out on uranium, carbide or nitride show however that for such compounds, a quantitative dissolution is accessible by using the conventional reagent of the Purex process (nitric acid), and that only minor adjustments are to be researched in order to optimize the operating conditions; The second point resides in the adjustment of the processes in order to enable a grouped management of the actinides: it involves researching the means of extracting all of the actinides (major and minor) from the dissolution solution in order to then develop the compound to be recycled; this includes the development of molecular architectures and appropriate process diagrams, in the continuation of the work carried out during the last decade on enhanced partitioning processes; the outline of such a concept, called GANEX, has recently been proposed by CEA (see Figure 80): in a preliminary stage, it is proposed to extract, the main part of the uranium contained in the spent fuel, then, in a sec- 69

68 ond stage, to partition jointly the plutonium and minor actinides (neptunium, americium, curium) by implementing an adapted version of the DIAMEX-SANEX process developed within the framework of the studies carried out pursuant to line 1 of the December 1991 law; an effort to integrate the recovery and remanufacturing operations for this grouped management of actinides to be recycled seems to be, in the same order of ideas, an orientation to retain; An important objective also resides in the development of the formulation of extracting agents, in order to increase their resistance regarding radiolysis phenomena; this would offer the possibility of reprocessing barely cooled fuel; Finally, the technologies and their implementation constitute a determining research avenue to increase the compactness of the processes, whether they concern single technologies (where remarkable progress margins have already been obtained with the development of liquid/low stay time liquid contactors) or their integration (advances in the field of online control seem to be important factors for the simplification of industrial workshop architecture). Irradiated fuel metals for the main part). Encouraged by the potential advantages of these concepts, exploratory studies, laboratory studies and technological developments have currently been initiated or revived by various research teams. The results which will be produced in the next few years will be essential in order to best identify their potential, best apprehend the difficult points and to orient the following phases of their development. There is still a long way to go before reaching the technical and industrial maturity of such processes. Independently from the process implemented, the strategic orientations retained for fuels of the future raise a certain number of questions, the relevance and intensity of which will depend finally on the options which will be fixed, but which already have to be considered at this stage. Below, nonexhaustively, a few examples: The concern for a close retention of fission products in reactor fuel, which leads to cladding or elaborated encapsulation devices being envisaged (particle fuels for example) may modify the accessibility of the materials to be recycled during reprocessing stages; new objects, new materials must be associated with suitable destructuration concepts; These matrix materials must obviously be managed: according to their abundance and the nature of the destructuration processes, their presence may increase the complexity of recycling operations; Dissolution Preliminary U partitioning Coextraction An + Ln FP Actinides to be recycled U Disextraction An Waste U + Pu + MA Disextraction Ln Ln The option of an integral recycling of actinides certainly leads to final waste with reduced toxicity. In counterpart, it entails a hotter recycled fuel, which will require remote operated remanufacturing processes; Particular attention is to be paid to the management of cycle effluents for some fuel options (carbon-14 with nitride fuels for example) or installation options (liquid discharges obviously entail more constraints for a recycling option on the reactor sites); Fig. 80. A grouped actinide extraction concept: GANEX. In the field of pyrochemical processes, the main objective of the research to be carried out resides in the confirmation of the potential of such concepts for industrial spent fuel recycling operations. Although absolutely significant developments and experiments have been carried out over a long period of time, few results currently concern the recovery of plutonium and, all the more so, minor actinides, and also the management of spent salts. A great number of avenues currently remain open, concerning both the choice of reactional media (fluorides or chlorides, but also room temperature ionic liquids, which are currently experiencing important growth), and that of technologies (electrolysis or extraction by molten According to some experts, one could also try to reduce the costs of final waste disposal by removing particularly thermogenous fission products (Caesium 137, Strontium 90, see Fig. 81) from the waste.this option adds some complexity to the processing operations, but could take advantage of the additional freedom that intermediate storage may provide. It deserves at least some studies, in the general framework of the optimization of the back-end of the fuel cycle. 70 The fuel cycle of future nuclear systems

69 Residual power (W/TWhe) 10, , Pu (reproc. at 4 years) Am (reproc. at 4 years) Cm (reproc. at 4 years) Fission product Cs Sr ,000 10, ,000 Cooling time (years) Fig. 81. Contribution of the various radionuclides with the residual power released by spent fuel (UOX, 55 GW.j/t). To summarize... The orientations outlined for the nuclear systems of the future give considerable importance to the fuel cycle operations (in particular regarding the range of materials to be recycled). We shall have to process new fuels, in a probably reinforced field of economical and environmental constraints. This multiple challenge calls for innovations and must be dealt with as a whole: the entire chain of reactors, fuel and cycle, must progress coherently. Two large avenues currently seem to be preferred: firstly hydrometallurgical processes, strengthened by consistent industrial feedback which attests their potential, and which still appear to have important margins of adaptation and progress; and then the pyrometallurgical processes, promising in some respects, but the potential of which are to be further explored. The magnitude of the field of research to be carried out shows all the interest of an organized international cooperation, such as is currently being established within the Generation IV Forum. Finally, as was reported during the forum s expert meetings, it is necessary to take into consideration the fact that the deployment of 4 th generation reactors may only intervene progressively, and that the 21 st century fleet will present a large component of water reactors, the cycle installations of which will also have to manage spent fuel, in order to produce final waste compliant with the specifications and criteria which will prevail, and in order to supply new generation reactors: this symbiotic character of the fleet will also constitute important input data for future choices. Nuclear energy of the future: what research for which objectives? 71

70

71 Uranium resources The element uranium Uranium is the heaviest of the natural elements remaining on Earth 24. Natural uranium mainly consists of two isotopes: U 235 and U 238. Isotope Period (years) Current relative abundance on earth (in % U total) millions billions This isotopic composition of natural uranium is found everywhere on Earth 25, no physical or chemical process having led to a significant separation of the two isotopes. Uranium enters into the composition of at least two hundred minerals, and its average content in the earth s crust is approximately 3 grams per tonne. It is present in practically all of the rocks of the Earth s crust, with particular concentrations in phosphates, certain igneous rocks or in the vicinity of oxidation-reduction boundaries in sedimentary rocks. Uranium is generally extracted from the subsurface by conventional hydrometallurgical and mining techniques. Uranium resources Today, most of the uranium produced in the world comes from Canada, followed by Australia and Nigeria. Large deposits of high-grade ore are yet to be to be exploited in Australia and Canada. Global reasonably assured resources (RAR), that can be recovered at a cost lower than $80/kg uranium, amount to approximately 2.5 million tonnes. Of course, resources depend on the price that one agrees to pay in order to recover them: thus, RAR resources that can be recovered for less than $130 per kg of uranium are estimated at 3.3 million tonnes. the RAR reserves currently estimated. Of these 2 million, only a part has been consumed in civilian reactors, leaving on the order of 1.2 million tonnes of depleted uranium with approximately 0.3% of U 235 which can be considered as a strategic reserve for the future. At the current rate of consumption (approximately 60,000 tonnes per year), cheap reserves should last between 50 and 100 years. Beyond this horizon, the millions of tonnes of uranium contained in phosphates and the billions of tonnes contained in the water of the oceans (the content is 3 parts per million) could be exploited. In reality, the future of the uranium resources will depend a great deal on the fuel cycle of the reactors which use them. Uranium is used quite unefficiently in water reactors: the extraction of approximately 200 g of natural uranium is necessary in order to obtain the fission of 1 gram of material in this type of reactor. If one continues to use uranium in open cycle light water reactors, the uranium reserves may seem modest in relation to those of fossil fuels. However, a single recycling, notion without significance for fossil fuels, already significantly increases the scope of the resources. In parallel, the use of fast neutron reactors would enable the energy potential of the uranium to be better used by efficiently consuming the fertile isotope U 238 in a closed fuel cycle 26. With these nuclear systems, the resources would no longer be a matter of concern. How vast are these reserves? By way of comparison, 2 million tonnes of uranium have been produced since the beginning of the nuclear power industry, that is a quantity close to 24. Tiny quantities of natural plutonium are found in uranium ore. This plutonium is formed by absorption of neutrons produced by the spontaneous fission of uranium, 25. With the exception of the Oklo deposit, where natural nuclear reactions took place which consumed uranium 235 and disrupted the isotopic composition of the remaining uranium. Nuclear energy of the future: what research for which objectives? Fig. 82. An open-cut uranium mine. 26. See supra and infra, pp. 67 and 75, the chapters entitled: The fuel cycle of future nuclear systems and The origin of species (of reactors). 73

72 Unit: Billion of toe Canada (13.0 %) Algeria 26.0 (1.8 %) Spain 3.1 (0.3 %) France 12.5 (0.5 %) Ukraine 42.6 (1.7 %) Russia (5.6 %) Kazakhstan (17.4 %) Mongolia 61.6 (2.5 %) United States (4.2 %) Nigeria 71.1 (2.8 %) Australia (24.2 %) Brazil (6.5 %) Other (4.9 %) World total: 2,506.2 billion tonnes (excluding Chile et China) Gabon 4.8 (0.2 %) Namibia (6.0 %) Fig. 83. Proven global reserves of uranium* ( ). * Reasonable resources assured recoverable for less than $ 80/kg U. South Africa (9.3 %) Source: Energy observatory at CEA/DES and IEA/OECD 74 Uranium resources

73 On the origin of species (of reactors): systems and generations Nuclear reactor design begins by the layout in the reactor core of fissile and fertile materials constituting the fuel, a coolant used to evacuate the heat produced by the fission reactions, a moderator (possibly) which slows down neutrons, and a neutron absorber to control the chain reaction. Several options are possible for each of these elements, and, while all of the combinations are not viable, many types of reactor can be envisaged. Fissile and fertile materials The most commonly used fissile nucleus in current reactors is U 235, a single natural fissile isotope. Other fissile nuclei that can be used are the odd plutonium isotopes Pu 239 and Pu 241, produced by neutron irradiation of the fertile isotope U 238. The mixture of fissile and fertile isotopes in the core enables the operating time of the core to be increased, since the disappearance of fissile nuclei by fission is partially compensated (or totally if the reactor is a breeder reactor*) by the formation of new fertile nuclei via neutron capture on the fertile nuclei. Coolant Many choices are possible for the coolant fluid: heavy water, ordinary water, gas (helium, CO 2 ), liquid metals, etc. The coolant may circulate directly from the core to the turbine or exchange heat with a secondary circuit. The choice of coolant has a great importance in the reactor s technology, and large systems are often classified according to it. Moderator «If any species does not become modified and improved in a corresponding degree with its competitors, it will soon be exterminated.» Charles DARWIN, On the Origin of Species by Means of Natural Selection, Another fundamental choice is that of the mean energy, or mean speed of the neutrons in the core. The choice between fast neutrons and slow neutrons thus determines two main groups: In slow neutron or thermal reactors, the neutrons are slowed down by successive collisions on the light nuclei of a moderator material. The main moderator materials used are ordinary water, heavy water (D 2 O) and graphite. As slow neutrons have large interaction probabilities with the material, this type of reactor may operate with a fuel little enriched with fissile nuclei (natural uranium may even possibly be sufficient), but only a small part of the energy from the fuel s heavy nuclei is exploited. A lot of these heavy nuclei are transmuted by neutron capture in actinides that will be found present in the waste; In fast neutron reactors (FR), the neutrons are not slowed down in the reactor, and they more or less keep the energy that they had during their production by fission. Their interaction probabilities with the material are low, this is why fast neutron reactors must have a high neutron flux, and contain a lot of fissile material. On the other hand, in this field of neutron energy, fission reactions are favoured in relation to parasite (capture) reactions: Fissile material is used much better than in a thermal neutron reactor. FRs are potential burners of actinides, the latter being fissile with fast neutrons. Cross section (barns) 1e6 1e5 1e4 1, e-4 1e-5 Fission Capture Slow neutrons Fast neutrons 1e-5 1e ,000 1e4 1e5 1e6 1e7 Neutron energy (ev) Fig. 84. The fissile and capture cross-section* of uranium 235 as a function of the energy of the neutron highlights two main fields: that of slow neutrons, where the interaction probabilities of the neutron with the uranium nuclei are high, and that of fast neutrons, where the cross-sections are much smaller. Nuclear energy of the future: what research for which objectives? 75

74 Current reactor families In the 1950s and 1960s, practically all nuclear reactor types were envisaged, designed and even built! Following this beehive of creativity, natural selection ensured the survival of a reduced number of families. Gas reactors Graphite-gas reactors enable the use of natural uranium. They developed in many countries (United Kingdom, France, Japan, Spain, Italy) until the United States, which up to the end of the 1950s maintained the monopoly on enrichment, agreed to export enriched uranium. From then on, all of these countries progressively abandoned this technology in order to switch to the light water reactor type. The last was the United Kingdom, which started its first water reactor in 1995, and is the only one to keep reactors of this type in operation today. RBMK graphite - water reactors It was this type of reactor which caused the Chernobyl accident. The graphite moderator is penetrated by zirconium alloy pressure tubes in which the boiling water circulates in order to cool the slightly enriched uranium fuel. This type of reactor is unstable in its design in certain operating domains, which makes it vulnerable to human error. The complete shutdown of this reactor family is programmed. Fig. 86. A RBMK reactor (unit no. 4 at Chernobyl). CANDU heavy water reactors Fig st to 2 nd generation reactors: the large NUGG reactor (shut down) and the small PWRs (in service) which succeeded it can be seen on the Bugey site. Ordinary water reactors With 86% of the fleet in operation and 79% of constructions in progress throughout the world, ordinary (or light ) water reactors represent the worldwide dominant species of nuclear reactors. PWRs, and their Soviet version, the VVER, are the most common. They are robust, reliable, and display continued progress in terms of availability, burnup rate, cycle time, ability to follow fluctuations of the power grid and collective dose to operators. Boiling water reactors (BWR), which represent approximately a third of the capacity of PWRs, have also seen significant development, but have been slightly hindered by a few teething defects. Today, in Japan, the last orders have exclusively concerned boiling reactors. In this type of reactor, the fuel is cooled by the circulation of heavy water in the pressure tubes. The heavy water moderator absorbs very little of the neutrons, which enables this type of reactor to use natural uranium. This specificity may attract countries wishing to free themselves of uranium enrichment. The Canadians exported CANDUs to many countries (India, Pakistan, Romania, Korea, China). The systems of tomorrow Even if water reactors are currently dominant, several types of reactor have specific advantages that may one day compete with them: High temperature reactors (HTR) HTRs are thermal neutron reactors, moderated by a large mass of graphite and cooled by helium circulation.they use an original fuel, the coated particle initially designed in England. This fuel, made from carbon and ceramic enables high refrac- 76 On the origin of species (of reactors): systems and generations

75 n tory cores to be made, operating at high temperatures, which offers the possibility of high efficiency thermodynamic cycles. The great freedom offered to the designer via the particle fuel makes this type of reactor suitable for accommodating a large variety of fuel cycles. Several HTR prototypes have been developed in the United States and in Germany. Made attractive by the recent progress in gas turbines, they are currently studied in the form of small modular reactors cooled by a helium circuit coupled directly to a turbine. With their large thermal inertia, HTRs are particularly safe, which may permit their safety systems to be simplified; their excellent thermodynamic efficiency should make it possible to amortize their investment cost rapidly, a cost which is still high due to their low power density. Fast neutron reactors (FR) The great advantage of fast neutron reactors resides in their ability to produce as much or more fissile material than they consume. Fast neutron breeder reactors may therefore, via successive recycling, use the quasi-totality of the energy contained in the uranium, one hundred times more than an ordinary water reactor. e U 238 U239 Np239 Pu min 2.35 Tage Fig. 87. Formation of a plutonium 239 (fissile) nucleus via capture of a neutron on uranium 238 (non-fissile). The fission of a nucleus produces several neutrons. Only one of these neutrons is necessary to maintain the chain reaction. The other neutrons may form other fissile nuclei via capture on uranium 238 in order to form plutonium 239. With a replenisher or breeder reactor, as much or more fissile material can be produced than consumed. The fissile material therefore plays the role of a catalyser, constantly regenerated during its consumption. With this type of reactor, it is the fertile material U 238, which is in fact finally consumed. e In thermal spectrum reactors, actinides often capture neutrons without fissioning, which leads to the formation of heavier and heavier nuclei, all radioactive, which put a strain on the neutron inventory of the reactor and is found in waste. In fast spectrum reactors, capture and fission coexist for all of the actinides, which offers the possibility of balancing their inventory. Still by way of comparison, a typical UOX-PWR (1GWe) 16 kg of minor actinides. The recycling of Pu in the MOX form enables the Pu inventory to be stabilized, but the minor actinides are not burned and build up. A FNR replenisher of the same power can consume the minor actinides that it produces 27 (see the block diagram of the FNR fuel cycle in chapter N). With this type of system, nuclear power may therefore gain in cleanliness. The only FRs on which we have significant feedback are the ones cooled by liquid sodium. This is an excellent coolant, not very corrosive for stainless steels when it is pure, but which spontaneously ignites with air and reacts quickly with water. The Russians are studying FR models cooled with molten lead, whereas the French are reopening, the helium-cooled FRs file after shutdown of the Superphénix 28. The investment cost of FRs is much higher than that of PWRs with the same capacity. FRs therefore only have a chance to emerge if or when their specific quality, fissile material economy, becomes a key factor of success. In a more distant future To complete the list of possible future reactors, it is finally necessary to mention the molten salt reactors and the ADS (Accelerator Driven Systems), hybrid reactors coupled with a proton accelerator. Nuclear technology is young, and there is no lack of ideas to adapt it to new global requirements in terms of energy and environment. What is certain, is that sustainable nuclear power will only exist within the framework of a responsible radioactive waste management and fissile and fertile material recycling strategy. By way of comparison, a typical UOX-PWR (1GWe) needs 110 t of natural uranium per year and produces 0.25 t of plutonium per year.a FNR replenisher of the same power would need 15 to 20 t of Pu (constantly regenerated), and would consume only approximately 1 to 2 tonnes of natural uranium per year. FNRs may even operate using the large stock of depleted uranium currently unused by the water reactor fleet. FNRs therefore solve the problem of resources. 27. See supra, p. 68, the block diagram of the fast reactor cycle. 28. In Creys-Malville (Isère). Nuclear energy of the future: what research for which objectives? 77

76 2040 Tomorrow 2005 Today 1986 Chernobyl st oil shock 1960 Blossoming of the reactor concept Gen.I Gen.II Gen.III Gen.IV WPu CANDU SGHWR (Steam generating) Heavy water reactors WPu Magnox (Graphite-gas reactor) WPu AGR VHTR HTR Graphite moderator reactors GFR RBMK (graphite moderator and water coolant) BWR (Boiling water reactor) U enriched reactors Naval propulsion Naval PWR + fuel reprocessing EPR Civilian PWR SCWR FR (Fast neutron reactor) FNR Na FNR Pb ADS MSR MSR (Molten salt) 1942 The beginning Fermi Joliot-Curie The nuclear reactor phylum Fig. 88. The phylogenetic tree of nuclear reactors. Brief description of the main branches of the tree: reactors may operate with natural uranium or enriched uranium, but the use of natural uranium restricts the choice of coolants to graphite and heavy water. The use of enriched U offers almost all possible choices of coolants and moderators. Some combinations are more fortunate than others: the water coolant has had a lot of success, because it is also a good moderator. Water reactors (PWR and BWR) constitute most of the contingency of generations II (current) and III (near future) reactors. The combination of a graphite moderator and a gas coolant paves the way to high temperature reactors. The branches of fast neutron reactors are still little developed. Only certain species of nuclear reactor have survived. Some branches are extinct or in the process of becoming extinct: NUGGs for economic competitiveness reasons, RBMKs for safety reasons. But the selection criteria are changing, the world is evolving. Other species are emerging. The six concepts retained by the Gen IV are at the top of the tree. Will they all be developed? WPu: Military plutonigenous reactor. SGHWR: Heavy water reactors supplying industrial heat (Steam Generating Heavy Water Reactor). AGR: Graphite-gas reactors (Advanced Gas-cooled Reactor). (V)HTR: (Very) High Temperature Reactor. SCWR: Super Critical Water Reactor. ADS: Hybrid spallation-fission system (Accelerator-Driven System). FR: Fast Reactor. MSR: (Molten Salt Reactor). 78 On the origin of species (of reactors): systems and generations

77 Systems of the future First creations Current reactors Advanced reactors Generation I UNGG Generation II CHOOZ PWR 900 PWR 1300 N4 Generation III EPR Generation IV Fig. 89. The nuclear generations calendar. Petite histoire des générations nucléaires The first generation of reactors saw the day when the industrial technology of uranium enrichment was not yet developed. Reactors had to be able to operate with natural uranium (nonenriched), hence the use of moderators absorbing very few neutrons, such as graphite or heavy water. This is why the field, called Natural Uranium Graphite Gas (NUGG), was developed in France. The second generation of reactors, deployed in the 70s to 90s, constituted most of the global fleet currently in operation. This period was that of pressurized water reactors PWRs and boiling water reactors BWRs. The cumulated operation of more than 10,000 years-reactors on the global level proves the industrial maturity and the economic competitiveness of this technology. The fleet of 58 pressurized water reactors that France has belongs to this second generation. The third generation represents the most advanced industrial state-of-the-art. It concerns socalled evolutionary reactors which benefit from the feedback and industrial maturity of second generation reactors, whilst integrating even more advanced specifications in terms of safety. Finally the development of the fourth generation has as from now been engaged, within an international framework and with the objective of bringing these new systems to technical maturity, with the prospect of industrial deployment by 2030.The purpose of these systems is to respond to the issues of sustainable energy production, with a long-term vision, and in particular, to minimise radioactive waste and to better use natural resources of fuel, as well as to meet new energy requirements: not only the generation of electricity, but also hydrogen for transportation and drinking water via the desalination of seawater. These systems present significant evolutions and technological innovations (they can be called revolutionary ) which require approximately twenty years of development Nuclear energy of the future: what research for which objectives? 79

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79 Nuclear systems of the future: an international framework for the development of a new generation of nuclear systems The Generation IV International Forum France United Kingdom The objectives targeted for the systems of the future, as well as the choice of key technologies to achieve them, are the subject of a very active international cooperation, in particular within the framework of the Generation IV Forum. Canada USA European Union Taking stock of the risks of shortages and medium-term energy dependence, the American government has committed itself to an effort to revive the means of generating electricity. In the field of nuclear energy, this results in two complementary actions: Brazil Argentina Members of the Generation IV international Switzerland Japan The first is purely American, and intends to facilitate the construction of new reactors in the United States in the short term (2010); it concerns the Nuclear power 2010 (NP 2010) programme. An ad hoc group, the Near Team Deployment Group (NTDG), has evaluated the reactors likely to be constructed between now and 2010, has identified the possible problems to be solved on the technical, regulatory or administrative level, and has proposed actions facilitating the shortterm deployment of these third generation nuclear reactors; The second is the Generation IV International Forum. Its founding principle is the recognition of the advantages of nuclear energy by the ten member countries. This energy could meet the growing energy needs throughout the world, in a procedure for sustainable development and prevention of the risks of climate change. This principle is recorded in the Forum s charter, and it is embodied by the commitment of an international R&D to define, develop and enable the deployment of 4 th generation nuclear systems by The member countries of the Generation IV International Forum are Argentina, Brazil, Canada, France, Japan, the Korean Republic 29, South Africa, Switzerland, the United Kingdom, Switzerland, the United States and the European Union. Other countries or international instances may also eventually join this research effort. 29. South Korea. Nuclear energy of the future: what research for which objectives? South Africa Methodology of the choice of technological orientations Three steps have already been taken: South Korea Fig. 90. Nuclear systems of the future: highly international R&D. The evaluation of designs proposed by the participating countries, according to a highly codified methodology (this task was carried out between April 2001 and April 2002); The selection of a small number of leading technological concepts judged as particularly promising during the evaluation (task carried out in May 2002); The elaboration of a development plan for these technologies, published in October 2002, preparing a later phase of international cooperation (main objective of the Forum from 2003). Straightaway, a clear agreement was affirmed among the participants on the main objectives of the Generation IV programme and on the procedure. Four main objectives ( goal areas ) were defined in order to characterize the systems of the future. They must be: Sustainable: this means saving natural resources and respecting the environment (by minimising the production of waste in terms of long-term radiotoxicity, and by optimally using natural fuel resources); 81

80 Economical: from the point of view of the investment cost per kwe installed, the fuel cost, the operating cost of the installation and, consequently, the production cost of the kwh, which must be competitive in relation to that of other energy sources; Safe and reliable: with ongoing research in relation to current reactors, and by eliminating the need to evacuate the population outside of the site as much as possible whatever the cause and the gravity of the accident inside the power plant; Resistant regarding the risks of proliferation, and easily protected against external aggressions. Approximately one hundred engineers and scientists have participated in the first phase of the Forum s work. For each system considered (e.g. water, gas, liquid metal reactors) technical groups have been made responsible for the evaluation of the various concepts proposed in relation to the objectives and criteria retained, and these groups were also responsible for the elaboration of R&D plans for the concepts finally selected. The evaluation methodology was developed and refined by a specific workgroup which reduced the four main progress objectives mentioned above to approximately thirty basic criteria. Transverse technical groups identified the necessary developments in the field of fuel, the cycle procedures, the materials, and the safety of the energy products for the various systems considered by the Forum. A coordination group led all of the Five basic criteria Saving natural resources Extracting the energy from fissile material efficiently Reducing the risk of proliferation Burning plutonium with an integrated fuel cycle Economy Safety Reduction of the production of waste Recycling and transmuting minor actinides Fig. 91. The criteria retained for the selection of nuclear systems of the future differ in their name and in their hierarchy from those retained for first and second generation reactors. Here, all of the criteria were placed on the table, and debated in the greatest transparency. They are all of purely civilian inspiration, and shared by the international community. The profitability and economy criteria of the resources (important for industrialists) remain important. More innovative, the safety, waste reduction (important for the public) and reduction of proliferation risks (important for politicians) criteria are explicitly mentioned. technical groups activity and assured the integration of the results in the documents of the various stages and in the final summary. The choices made within the Forum Six nuclear systems were selected, which may enable notable advances on the abovementioned criteria. These systems enable applications other than electricity production, such as the generation of hydrogen or seawater desalination. The diversity of needs to be covered and the international contexts explain that we do not end up with only a Generation IV system, but with a few of the most promising system designs, on which the Forum R&D member countries are now concentrating. Identity cards of the selected systems The selection operated in the Generation IV initiative shows several important lessons: In the choices retained the most discriminating criteria were those of sustainable development. The range of evaluations on the economical or safety aspects turned out to be much more narrow. This results in a majority of fast spectrum and closed cycle systems; The most innovative concepts find themselves penalised by great uncertainties regarding their definition and regarding the possibility of removing technological difficulties for a production between now and In this class of nuclear systems, the final choice falls on the molten salt reactor, interesting for the management of actinides and the deployment of the thorium cycle; The grouping into groups of reactors homogenous from the performances and R&D requirements point of view proved to be important because it enabled the R&D knowledge bases to be taken into account and recommendations around important federal policies to be structured. By way of example, the gas-cooled group of reactors (RCG), comprizes an important research knowledge base regarding high temperature materials, helium circuits, and conversion by gas turbine. In addition, different variants are being studied for various market niches: very high temperature reactors for the mass production of hydrogen, specialized reactors for burning actinides, fast neutron spectrum version and integral recycling for sustainable energy development; The various gas reactors (GFR, VHTR) translate the recognition of the interest for this coolant with, in particular, the possibility that it offers for developing an upgradeable range of systems based on this technology. 82 Nuclear systems of the future: an international framework for the development of a new generation of nuclear systems

81 SFR: improved sodium Pump Cold plenum Hot plenum Control rods Primary sodium (Hot) Core Heat exchanger Fig. 92. SFR: improved sodium. This system includes a fast spectrum reactor associated with a closed cycle enabling the recycling of all of the actinides and the regeneration of plutonium. Due to the regeneration of fissile Pump Primary sodium (Cold) Steam generator Pump Turbine Secondary sodium Condenser Generator Heat sink Electrical power material in the core, this type of reactor may operate for a very long time without intervention on the reactor core. Two main options are envisaged: the first, associated with a reprocessing of metal fuel, leads to a reactor with an intermediate unit power of MWe the second, characterized by a reprocessing of mixed oxide fuel (MOX), corresponds to a reactor with high unit power, between 500 and 1,200 MWe (reactor associated with PUREX reprocessing). The SFR presents very good natural resources usage and actinide management properties. It was evaluated as having good safety characteristics. The oxide fuel system may be ready for industrial deployment as of Several SFR prototypes exist throughout the world, in Japan (Joyo, Monju), Russia (BN600), and France (Phénix). The main research issues concern the integral recycling of actinides (fuels comprizing actinides are radioactive, therefore complicated to manufacture); the in service inspection (sodium is not transparent); safety (passive safety procedures are being studied); the reduction of the investment cost (this type of reactor is still expensive). The changing of the water of the secondary fluid for supercritical CO 2 is also being studied, because it may enable the safety to be improved, whilst allowing the elimination of the intermediate sodium circuit, if the chemical sodium-co 2 interactions proves to be less violent than sodium-water interactions. Fig. 93. LFR: a lead concept. This system comprizes a fast neutron reactor associated with a closed fuel cycle, enabling optimal use of the uranium. Several benchmark systems have been maintained in the selection. The unit powers go from MWe, for the so-called battery concepts, up to 1,200 MWe, including the modular concepts from MWe. The battery concepts have a long-term fuel management (10 to 30 years). The fuels may be either metal, or of the nitride type, and enable the recycling of all actinides. The main technological deadlock of the system concerns corrosion by liquid lead. LFR: a lead concept Header U-tube heat exchanger modules Reactor module/fuel cartridge (removable) Coolant module Coolant Reactor core Generator Turbine Compressor Heat sink Intercooler Compressor Pre-cooler Electrical power Recuperator Reactor Nuclear energy of the future: what research for which objectives? 83

82 Fig. 94. SCWR: water, but supercritical Two fuel cycles are envisaged for the SCWR, which correspond to two different versions of the system: a thermal spectrum reactor associated with an open fuel cycle and a fast spectrum reactor combined with a closed cycle for recycling all the actinides. Both options have an identical operating point in supercritical water: pressure of 25 MPa and core outlet temperature of 550 C enabling a thermodynamic efficiency of 44%. The unit power of the benchmark system is 1,700 MWe. The SCWR was evaluated as having high potential for economic competitiveness. The main research issue concerns corrosion by water, in particular accelerated in relation to current water reactors due to a much higher operating temperature. SCWR: water, but supercritical Control rods Reactor core Reactor Supercritical water Turbine Condenser Generator Electrical power Heat sink Pump VHTR: making hydrogen with helium? Control rods Reactor Graphite reactor core Graphite reflector Blower Helium coolant Pump Heat exchanger Heat sink Hydrogen production plant Water Oxygen Hydrogen Fig. 95. VHTR: making hydrogen with helium? The VHTR is a gas-cooled system associated with a thermal spectrum core and an open fuel cycle. The particularity of the VHTR is its operation at very high temperatures (>1,000 C) to supply the necessary heat for water decomposition processes by thermal chemical cycle (iodine/sulphur) or high temperature electrolysis. The VHTR is dedicated specifically to the production of hydrogen, even if it must also enable the generation of electricity (alone or in co-generation). The benchmark system has a unit power of 600 MWth and uses helium as a coolant. The core is made up of prismatic blocks or pebbles. Important research topics for the development of this system concern high temperature materials and the development of hydrogen mass production technologies. 84 Nuclear systems of the future: an international framework for the development of a new generation of nuclear systems

83 Fig. 96. GFR: fast gas The GFR is a fast spectrum system enabling the homogenous recycling of actinides whilst maintaining a regeneration gain greater than 1. The benchmark concept is a helium-cooled once through reactor with high efficiency (48%). The evacuation of the residual power in the event of depressurisation implements natural convection. The power density in the core is determined in order to limit the transient temperature of the fuel to 1,600 C. The innovative fuel is designed to retain fission products (for a temperature lower than the limit of 1,600 C), and to avoid their release in accident situations. The recycling of spent fuel is envisaged on the same site as the reactor either via a pyrochemical process or via a hydrometallurgical process. The GFR is the most performant concept in terms of natural resource usage and reduction of long-term waste. It is located in the technological gas line, complementing thermal spectrum concepts, GT-MHR 30, PBMR 31 and VHTR. The main research topics associated to the development of the GFR concern the reactor materials, which must be able to resist both high temperatures and strong neutron irradiations. The most important issue is the development of a dense and refractory fuel. GFR: fast gas Reactor core Control rods Reactor Helium Heat sink Generator Turbine Intercooler Compressor Pre-cooler Electrical power Recuperator Heat sink MSR: a 2 in 1 system Reactor Purified salt Chemical processing plant Control rods Freeze plug Emergency dump tanks Coolant salt Pump Pump Generator Heat exchanger Turbine Intercooler Pre-cooler Compressor Electrical power Recuperator Heat sink Fig. 97. MSR: a 2 in 1 system The MSR is an epithermal spectrum system with the highly original implementation of a molten salt solution used both as fuel (liquid) and coolant. The regeneration of the fissile material is possible with an optional uranium-thorium cycle. The MSR integrates in its design an online recycling of fuel, and thus offers the opportunity of grouping on the same site electricity generating reactor and its reprocessing plant. The salt retained for the benchmark concepts (unit power of 1,000 MWe) is a sodium, zirconium and actinide fluoride. The spectrum moderation is obtained in the core by the presence of graphite blocks crossed by the fuel salt. The MSR comprizes an intermediate fluoride salt circuit and a tertiary water or helium circuit for the generation of electricity. This system was evaluated as having relatively good safety and non-proliferation characteristics. The most important research issue concerns the development of the online molten salt fuel recycling technology. 30. GT-MHR : Gas-Turbine Modular High Temperature Reactor. 31. PBMR : Pebble Bed Modular Reactor. Nuclear energy of the future: what research for which objectives? 85

84 The second stage of work from the Forum is the international co-operation phase to consolidate systems feasibility by removing technological deadlocks and validating their performances. It is currently being established and France is playing a very active role. The systems, whose feasibility is to be confirmed, will enter into a validation phase of their technical and economic performances. According to the degree of innovation of the system, all this work should lead to sufficient technical maturity between 2015 and It should enable important industrial deployments by What research for the nuclear systems of the future? Research on nuclear systems of the future must be based on a quality modelling. The basic physical phenomena are mostly well-known, which does not signify that their modelling is easy! Fortunately, the progress of computer tools enables ambitious modelling to be envisaged. A new generation of calculation codes is in the process of development in order to describe the behaviour of nuclear systems: these software platforms use a multiscale (from microscopy to macroscopy) and multidisciplinary (taking into account the interactions between neutronics and thermohydraulics, for example) approach. In reactors of the future, the materials in general and the fuel in particular will be subjected to severe conditions, due to the high temperatures envisaged in certain reactor concepts, and International initiatives complementing the Generation IV International Forum INPRO In 2000, the International Atomic Energy Agency (IAEA) launched the INPRO project (International Project on Innovative Nuclear Reactors and Fuel Cycles), which aims to promote the development of innovative nuclear systems enabling future energy requirements to be met whilst respecting the objectives of economic competitiveness, safety, respect for the environment, resistance to proliferation, and acceptance by the public. The importance of this project is to accompany and complement technological developments, such as those carried out within the framework of the Generation IV Forum, there where IAEA may have a specific contribution, for example by enabling the participation of numerous countries, in particular developing countries not yet using nuclear energy but interested in benefiting from it, or thanks to its effectiveness in non-proliferation and international controls. Firstly (phase 1), the technical objectives of the project are: To determine, over a very large basis, the needs and objectives of countries, given the diversity of their situation, and to specify how innovative nuclear systems may contribute to meeting them; To define the criteria and methodologies for the analysis and comparison of various innovative reactor concepts. Secondly (phase 2), the Agency envisages that the project may extend the definition of the criteria and the evaluation methodology in order to help member countries of the Agency in their own analysis of nuclear systems that best meet their needs. Different from the Generation IV Forum, the purpose of the project is not to carry out technical R&D actions or to develop innovative reactors and systems. The European MICANET and HTR-TN networks The objective of the European MICANET network (MICHELANGELO Network) is to develop a European R&D strategy in the field of innovative systems and to contribute in defining projects from the 6th European R&D Framework Programme in relation to the Generation IV Forum s activity to enable exchanges best serving the interests of European players. The HTR-TN network is more specifically dedicated to gascooled systems. Bilateral cooperations The bilateral cooperation actions with the United States, Japan and Russia were redefined in 2001 with the aim of preserving a growing place for joint studies and developments regarding the gas reactor technology, the extrapolation of this technology to fast neutrons, and the development of fuel reprocessing and remanufacturing processes, with integral recycling of actinides. The cooperation with the United States carried out since 2002 to work with five common joint-financed projects on these topics (NERI-International actions within the framework of the CEA-DOE cooperation). Eventually, four of these projects may integrate Generation IV cooperation. The cooperation with JNC (Japan Nuclear Cycle Development Institute) enables the comparison between the gas-cooled and sodium-cooled fast neutron reactors to be extended, as well as the sharing with JAERI of certain technological developments (fuels, materials) and experimentation possibilities on their experimental helium-cooled HTTR reactor. 86 Nuclear systems of the future: an international framework for the development of a new generation of nuclear systems

85 caused by irradiation by the high flux of fast neutrons. Corrosion is in general accelerated at high temperatures, and this topic represents a research subject in itself. Irradiation damages caused in the materials by fast neutrons are qualitatively different from those caused by slow neutrons, because of the possibility that the former have of producing nuclear reactions. Refractory alloys and ceramics, solid or composite, are good candidates for nuclear applications. These materials have recently made spectacular progress and are applied in numerous industrial fields, but their adaptation to nuclear needs will require work. One of the major barriers for the development of nuclear systems of the future is the fuel itself, which must combine mechanical and thermal resistance characteristics under irradiation, whilst complying with the constraints linked to neutronics which severely restrict the geometry and the materials that can be used. For example, one of the greatest challenges in the production of a gas-cooled fast reactor will be to design a dense and refractory fuel. Gen IV concepts are not only nuclear reactors: they are designed to operate with a well determined fuel cycle. The reprocessing-recycling of fuel depends a great deal on the nature of the fuel, and on the reactor that can consume it.this is why we do not speak of an isolated reactor, but rather system to encompass the reactor and the reprocessing-recycling of its fuel. Consequently, the partitioning, storage and transmutation of nuclear materials involved in these cycles will remain important research topics. Physical metallurgy Atoms Atomic clusters Discrete Dislocation Dynamics Dislocations Digital mesoscope Grain Mechanical metallurgy Fig. 98. Example of multiscale simulation, applied to materials U Pu Recycling of the Pu in the LWR (MOX) Gen. II Pu (U) U Recycling of the Pu and MA of the LWRs in Gen IV fast reactors Polycristalline aggregate Ganex on spent LWR fuel (MOX et UOX) Gen. III Sodium-cooled reactors: an expertise which remains on the agenda The objective of maintaining and upgrading the expertise is applicable in particular to sodium-cooled fast reactors, on which France has acquired a Local approach of the fracture Materials R.E.V. (representative elementary volume) U, Pu, AM Structure mechanics Gen. IV U, Pu, AM Global recycling of the actinides in Gen IV fast reactor Fig. 99. The succession of fuel cycles associated with the generations of reactors. Currently, the plutonium from PWRs is recycled in MOX form. In 2020, generation IV PWRs will continue to exist, but the Pu that they produce will be burned (partially, but more efficiently) by the generation III reactors deployed at this time. The minor actinides produced by this mixed Gen II - Gen III fleet may be partitioned and stored. In 2040, the first generation IV reactors will be deployed, and will burn the Pu which will have been placed in reserve for their start-up, in addition to the minor actinides accumulated earlier. The uranium complement necessary for the operation of these reactors may be supplied by currently stored depleted uranium. Around 2050, these Gen IV reactors should be able to operate by recycling the totality of their actinides. major technological advance in terms of R&D, experimentation and industrial developments. Thanks to the knowledge acquired during the development of the Phénix and Superphénix reactors and the EFR project, CEA masters all of the aspects of the sodium-cooled fast reactor system: The creation of installations since the experimental Rapsodie reactor (40 MWth) up to the industrial Phénix (563 MWth) and Superphénix (3,000 MWth) prototypes; Nuclear energy of the future: what research for which objectives? 87

86 Industrial mastery of the main stages of the fuel cycle: - manufacturing of uranium and plutonium based fuels, - reprocessing of spent fuel with a 1981demonstration of Phénix s ability to use the plutonium that it produced itself during a previous cycle; Experience of the good in service behaviour of a large range of structure materials (mainly steels). Such an experience is used within the framework of the research on waste management, because the Phénix reactor is currently used successfully for a series of experiments on actinide transmutation. This expertise is also upgraded via international cooperation, mainly with Japan and the United States, within the Generation IV Forum, as well as with Russia. One of the main challenges of this jointly carried out research is to provide sodium-cooled FRs with a good level of economic competitiveness, by making them more compact, and therefore cheaper on investment. CEA is also working on the SMFR, sodium-cooled modular fast reactor concept with the Argonne laboratory and the Japanese research institute JNC. This reactor has the particularity of modest power and a very long stay time of the fuel in the reactor. One of the developments that can be envisaged for sodiumcooled reactors consists of replacing the water of the secondary circuit by another fluid less likely to react chemically with sodium. The case of a secondary circuit using supercritical CO 2 is currently being explored in detail at CEA. Would we know how to make an FNR-Na with a secondary circuit using supercritical CO 2? What would be its advantages and its disadvantages in relation to a secondary circuit with water, in terms of safety, and efficiency? Gas-cooled reactors (GCR): a preferred development point Within the framework of the Generation IV International Forum, France has expressed a preferential interest for advanced very high temperature gas-cooled (VHTR) systems and for fast neutron systems with integral actinide recycling (GFR). It will also accompany developments regarding the fast neutron and sodium-cooled system (SFR). The very good positioning of gas reactors in the final evaluation, and therefore the recognition of the interest for this concept by the Generation IV Forum, backs up the decision made by CEA in 2000 to extend its research on this topic. Fig Reactor hall of the Phénix power plant. Established on the edge of the Rhône, an integral part of the Marcoule nuclear site, Phénix is a sodium fast neutron reactor. Its first divergence* took place in 1973 and the first kilowatt-hours delivered on the grid in July The last few years have been marked by important renovation works. The experimental programme mainly concerns actinide transmutation, but the experience acquired also benefits the research on nuclear systems of the future. 88 Nuclear systems of the future: an international framework for the development of a new generation of nuclear systems

87 Gas-cooled reactors Gas-cooled reactors are currently experiencing renewed interest due to their high operating temperature, which enables a high efficiency energy conversion cycle, and nuclear energy uses other than the generation of electricity. In their thermal spectrum version, construction of industrialscale reactors are possible in the medium-term. These reactors present recognized safety characteristics, as well as a great deal of flexibility in the choice of fuel cycle.this is permitted via the association of three essential specificities: a particularly confining particle fuel, a coolant, chemically inert helium (He), and finally, the exceptional physical properties of graphite as a moderator and structure material. In their fast spectrum version, which is still on the drawing board, they offer the additional prospects of energy upgrade of natural uranium resources, within the framework of the fuel cycle reducing final waste and the risk of proliferation. The relevance of this choice as a main avenue of research and development has been validated by the member countries of the Generation IV International Forum, who have retained two of the systems proposed by CEA (the VHTR and the GFR), from the most promising progress concepts for the next few decades 32. Thermal spectrum gas-cooled reactors The thermal spectrum High Temperature Reactor (HTR) concept differs notably from the other gas-cooled thermal neutron reactors which have been developed in the past: MAGNOX and AGR in Great Britain, and NUGG in France. In relation to these concepts, HTRs differ by: The use of the He coolant enabling access to high temperatures ( 850 C), hence a much greater thermodynamic efficiency; The use of a finely divided fuel made up of coated particles which gives it much higher burnup rate capacities, and opens the possibility of using different nuclear matter. HTRs and other main systems HTR BWR PWR FNR Unit power 200-1,000 1,100 1,450 1,200 type (MWe) Efficiency (%) Coolant He eau eau Na Pressure (bar) Inlet T ( C) Outlet T ( C) Moderator graphite water water without Power density (MW/m 3 ) Burn-up rate (GWd/t) The most recent ideas of modular design for HTRs further strengthen their attractiveness from the point of view of safety, economy and the possibilities of deployment. The use of gas turbines finally enables a once through energy conversion cycle (Brayton cycle), to be envisaged, improving the efficiency and the compactness of the installation. These are the reasons which contribute to the renewed interest in this system. Particle fuel The progress carried out in industry on gas turbines and high temperature materials has paved the way for HTRs with once through cycles, offering new prospects for increasing the thermodynamic efficiency of the energy conversion system. In addition, significant advances in the technology of heat exchangers and magnetic bearings currently enable more compact, cleaner and safer gas power plants to be designed. 32. The DEN no. 1 monograph (to be published in 2006) will be entirely dedicated to gas-cooled reactors. All of these elements are originally modular HTR concepts, which illustrate the industrial projects, such as the GT-MHR designed by General Atomics, the PBMR developed by Eskom in South Africa or the Antares project from Framatome-ANP. Nuclear energy of the future: what research for which objectives? 89

88 Current trends for the HTR system are therefore to be considered: Modular reactors with unit power in the 100 to 300 MWe range; Operating in once through cycle according to the Brayton cycle; Enabling the evacuation of the residual power to be assured passively and without using coolant fluid. The very high temperature reactor (VHTR) Fig The use of particle fuel constitutes the main innovation of HTRs. The kernel of fissile material (UO 2, PuO 2, UC, etc.) is surrounded by several successive layers (porous or dense pyrocarbon, SiC), used to assure the protection of the fissile kernel, and the containment of fission products. The whole thing is refractory (not metal) and highly resistant, which enables this fuel to be forced to very high temperatures and very high burnup rates. This fuel has already been used successfully in the past. It is still capable of performance progress, by a judicious choice of coating materials (all have not been explored, in particular, the replacement of the SiC layer by ZrC paves the way to very high temperatures, in the order of 1,000 C). CEA is currently equipping itself with a pilotinstallation for manufacturing this type of particle fuel (GAÏA installation, in Cadarache). Beyond the medium-term future, mentioned above regarding the HTR, the gas-cooled reactor system has the ability to develop towards even higher temperatures, with at stake a considerably improved energy conversion efficiency. Enhancement of nuclear heat utilization Light water ractor Steam temperature: 300 C Steam cycle Thermal efficiency 35 % Power generation 300 C Loss 30 C High temperature gas-cooled reactor Gas temperature: 1,000 C 950 C Cogeneration 600 C Hydrogen production Loss Thermal efficiency 70 % Power generation Process heat 30 C 250 C Fig Thermal energy is converted a lot better, if it is produced at high temperatures. With a PWR: 2 GWth are discharged to produce 1 GWe; With a VHTR: only 1 GWth would be discharged to produce the same electric power. But this type of reactor also enables the cogeneration of hydrogen and industrial heat to be carried out, which may bring the global conversion efficiency to approximately 70%. Fig The ANTARES project from Framatome-ANP has a capacity of 600 MWth. It uses helium at 850/1,000 C with an intermediate heat exchanger, and has a wide range of applications. In addition, high temperatures pave the way for other nuclear energy industrial applications, in particular the production of hydrogen. 90 Gas-cooled reactors

89 VHTR Transmission of electricity Heat CO 2 Electricity Electrolysis Hydrogen Thermochemical cycle, e.g. lodine/sulphur Methane reforming Industrial storage Distribution H 2 O Energy, heat Hydrogen vehicle CO2 2 CO 2 sequestration Fig Future nuclear systems may produce both electricity and hydrogen. To produce hydrogen from nuclear energy? The preoccupations linked to climate change, combined with important progress carried out recently on fuel cells, makes the use of hydrogen as a clean energy vector more interesting than ever. The American government has identified hydrogen as an essential element for the future economy, both for industrial needs such as the hydrogenation of heavy oils into light fuels, or as transportation fuel. However, hydrogen is not a primary energy and must be produced, either by electrolysis or thermochemical dissociation of water. The high temperatures which may be reached in nuclear reactors position gas-cooled reactors remarkably well for hydrogen mass production applications. Currently, the very large majority of the global production of hydrogen comes from reforming natural gas: Q+ CH H 2 O CO H 2, which produces a lot of CO 2, both in the chemical reaction and in the calorific contribution to the endothermic reaction. (and price) that such a production would entail, the abovementioned schema does not solve the problem of pollution. It is recognized that the only hydrogen mass production methods emitting no greenhouse gas are the high temperature electrolysis and the thermochemical partitioning of water from electricity and nuclear heat. The production of hydrogen by thermochemical method may be carried out in many ways. One of the ways preferred by numerous research laboratories is the I-S process, thus baptized because it makes the two reagents, iodine and sulphur intervene (without consuming them). The process involves the decomposition of sulphuric acid, a stage at which the efficiency is highly dependent on the temperature. A desired efficiency of 50% requires a temperature of 900 C on the process level, that is approximately 1,000 C for the coolant exiting the core. In practice, only a gas-cooled reactor has the possibility of meeting this requirement. It is from the latter that the main characteristics of the very high temperature reactor follow. The United States projects a quadrupling of its hydrogen consumption between now and 2017 to 10 million tonnes per year. Clearly, apart from the strong increase in gas consumption Nuclear energy of the future: what research for which objectives? 91

90 O 2 H 2 T~850 C H 2 SO 4 Decomposition HI Decomposition T ~ 450 C Materials Heat H 2 SO 4 Distillation H 2 O HI Distillation It involves finding and developing materials able to resist both high neutron fluence and high temperatures. The research concerns refractory alloys, ceramics, and cermet and cercer composites. T~120 C Bunsen reaction 16H 2 O+9I 2 +SO 2 (H 2 SO 4 +4H 2 O)+(2HI+10H 2 O+8I 2 ) Helium technology, components, equipment Fig Diagram of the thermochemical cycle of hydrogen production via the iodine-sulphur process. What research for the VHTR? The development of the VHTR will not be easy. Admittedly, the German AVR reactor has already reached core outlet temperatures greater than 950 C. But above this temperature, technological ruptures become necessary. The main avenues of research are listed hereafter: Calculation tools and methods: HTR or VHTR cores present both a random geometry and heterogeneities on very diverse size scales. These two characteristics require an adaptation of the neutron calculation tools: one of the channels pursued is the development of Monte Carlo type neutron calculation methods. On the other hand, in gas reactors with once through cycle, the thermohydraulics of the core is strongly coupled with that of the turbomachine: any change in the operation of one has repercussions on the operation of the other. These couplings must be taken into account and modelled in order to assure control of the thermohydraulic behaviour of the system. Fuel technology The fuel is one of the deadlocks of gas reactors. For the VHTR, above all it involves finding a refractory fuel. Even if they remain yet to be qualified, solutions already exist with UCO for the fuel and ZrC for the cladding material. First generation gas reactors use CO 2 as a coolant fluid, and helium has been little used in nuclear power. Current research concerns tribology in helium, gas purification techniques, heat exchangers, pumps, turbines, as well as thermodynamic schemas enabling the best energy efficiency to be obtained from a helium-cooled reactor. Helium technological loop (Helite) Materials Helium technology Component tests Equipment tests Heating 1,000 C 100 C Test section 500 C 500 C Cooler Test section 1,000 C HPC 1,000 C 600 C Recuperator 200 C Compressor (1 MW, Q 0,4 kg / s, T< 950 C, P > 7MPa) Cooler Fig Example of research carried out for the development of the VHTR: CEA develops the test benches and loops for testing the main components and equipment associated with helium technology. 92 Gas-cooled reactors

91 The gas-cooled fast reactor (GFR) In a long-term perspective aiming to meet the sustainability requirement, the Gen IV forum has retained the gas-cooled fast reactor as a particularly interesting system.the latter must succeed in conciliating both the advantages of high temperature gas reactors with those, known, of fast neutron reactors (optimal use of resources, reduction of waste production, transmutation of actinides). In addition, the specifications of the integrated cycle would reduce the risks of proliferation. The reactors proposed will be based on the helium technology developed for the HTR and VHTR projects. Their specificities are the fuel and its cycle, and the safety of the reactor. Their fuel cycle is at odds with the existing one because it is proposed to not partition U and Pu, and also to not partition major actinides (U, Pu) from minor actinides (Np, Am, Cm). The core s design (without blanket) will target the iso-generation of plutonium and a non-proliferating cycle maintained only by the provision of depleted uranium. The first studies have enabled the image of fuel for a gas fast reactor to be outlined. The latter must combine a high density of fissile material with a good resistance to high temperatures and to irradiation by fast neutrons. Several fuel concepts are currently being studied at CEA: a dispersed fuel, in which the fissile compound is presented in the form of grains or millimetric sticks dispersed within a containing matrix assuring the function of 1 st barrier just like the PyC/SiC coatings of the HTR particle; a fuel rod type concept with leaktight ceramic cladding is also being evaluated. Towards a European demonstrator of 4 th generation gas-cooled reactor With the High Temperature Engineering Test Reactor (HTTR, 30 MWth) which has been exploited since 1998 by JAERI, Japan currently has the most efficient test means on very high temperature nuclear technologies, and on the nuclear production of hydrogen. The United States is preparing a first demonstration of the production of hydrogen by thermochemical decomposition or electrochemical decomposition of water in the Next Generation Nuclear Plant (NGNP, 600 MWth) project on the national Laboratory site at Idaho. South Korea and China are mentioning similar projects around > Fast neutrons > Integral recycling of the actinides GFR HTR R&D Particle fuel Materials He circuit technology Calculation systems Fuel cycle VHTR R&D VHT resistant materials Intermediate heat exchanger ZrC coated fuel H 2 production R&D Fuel for fast neutron core Cycle processes Safety systems Fig High temperature gas reactors are already relatively mature, and may be deployed in 3+ Generation. These reactors may then develop towards even higher temperatures (VHTR), and/or towards a fast spectrum (GFR). Nuclear energy of the future: what research for which objectives? 93

92 In this context, and given the industrial issues associated with the very high temperature and nuclear production of hydrogen technologies, CEA is studying a Study and Technological Development Reactor (REDT) at the Cadarache Centre, aiming to demonstrate the technological principles of the GFR. The main development stages of the REDT may be: A first operating phase at C with a thermal neutron core aiming to demonstrate the European mastery of updated HTR technologies and of high temperature heat conversion processes for various applications: electricity, production of hydrogen by thermochemical cycle or high temperature electrolysis. Other applications may eventually complete the range of demonstrations possible: gasification of the biomass or desalination of seawater; A second operating phase, around 2020, with a fast neutron core aiming to demonstrate the operating principles and specific technologies of the GFR (the fuel in particular), thus repeating the initial objectives of the REDT. The REDT enables the upgrade of the wide experience acquired in Europe on the high temperature reactor systems, and strengthens the position of European industrialists in the international competition to market reactors from this system around The REDT could be considered as the main element of a European test platform for key technologies for the VHTR and the GFR, as well as for the various high temperature heat applications. The purpose of this platform would be to supply the necessary experimental conditions for the studies regarding the high thermodynamic efficiency cycles for the generation of electricity and the development of cogeneration processes. The platform would also enable the qualification of components for the conversion of energy (turbines, heat exchangers), and the study of processes for the production of hydrogen, gasification of the biomass and desalination of seawater. The pre-project studies for this platform would come into the framework of the 7 th European research and development framework programme as of Gas-cooled reactors

93 Other avenues for the distant future: thorium cycle, hybrid systems, fusion The thorium cycle Thorium (Th 232) is a fertile material which is abundant in nature. By absorbing a neutron and then by radioactive decay, it produces Pa 233 then U 233, a fissile isotope. The latter is interesting, because its fission generates slightly more neutrons than that of U 235 or Pu 239 in a thermal spectrum. In the 50s, these different reasons have led to interest in the U 233-thorium cycle; fuels were manufactured and used in various reactors, of which the American Shippingport experimental PWR, the Fort St. Vrain HTR and the German THTR. Unfortunately, the emission of high energy (2.6 MeV) γ radiation by Tl 208 formed in recycled U 233-thorium fuels poses serious radiation protection problems in fuel manufacturing installations; this disadvantage is one of the reasons why the uranium-plutonium system is preferred 33 (the main reason being that it would be necessary in any case to start a thorium system with the only fissile material existing in nature, U 235; the thorium system, as opposed to the uranium system, may therefore not be developed alone). During the last few years, the thorium system has been the subject of a new study, both because this system produces much fewer transuranium elements* and because robotics and remote handling have made considerable progress, perhaps limiting the disadvantages linked with γ radiation. The results of these studies are summarized below: The best use of thorium is in molten salt thermal neutron reactors, which allow a reduced inventory in fissile material and which are just as favourable on the resources level as on the waste level (reduction of the production of U 232 source of Tl 208, of the reprocessing losses, of the consequences of accidental discharges, of the final waste); however it does not enable doing without U 235 or Pu in order to start the cycle and thus does not completely eliminate minor actinides; A Th-Pu cycle in a fast neutron reactor (critical or sub-critical) enables twice as much plutonium to be consumed as a U- Pu cycle (thanks to the absence of U 238), and large quantities of U 233 to be produced; once started, the U 233-thorium cycle may be self-sustained; 33. This disadvantage only exists in the manufacturing of solid fuel; it is drowned in the highly radioactive background of a reprocessing installation integrated near to a molten salt reactor. Nuclear energy of the future: what research for which objectives? A serious doubt remains on the possibility of using highly enriched U 233; if an enrichment greater than 20% were prohibited for non-proliferation reasons, a significant quantity of actinides would be found in the uranium-thorium cycle; The long-term radiotoxicity (1,000 years and beyond) of waste is dominated by residual U 233 and by several radionuclides: Pa 231, U 232, U 234, Np 237. In most of the cases studied, beyond 10 4 to 10 5 years, uranium-thorium cycles lead to a radiotoxic inventory which may be much higher than that of uranium-plutonium cycles. At that time, however, the radiotoxicity will have greatly decayed; A fast U 233-thorium reactor would be a good incinerator of minor actinides, but the benefits from the point of view of radiotoxic inventory of buried waste would not be significant beyond 10 5 years; Conversely, the heat release from actinides produced in thorium-based cycles is much lower than in uranium-based cycles; the result of this is that the thermal dimensioning of the disposal is only defined by the residual power of the fission products. By contrast, the uranium-plutonium fuel cycle is handicapped by actinides with high thermal release (curium and, to a lesser extent, americium); Once the thorium is extracted from the mine, the daughter radionuclides which remain in the mine tailings decay very quickly, at the rate of the period of 5.7 years from their top series, Ra 228; it follows that, as opposed to uranium mine tailings, the thorium mine tailings do not pose a real longterm problem. The thorium based systems are therefore similar to uranium systems, as regards fission products and the quantities of actinides in the very long-term; they are important for the thermal dimensioning of disposal, but present certain disadvantages for the remanufacturing of solid fuels after reprocessing (but the problem is the same for GEN IV cycles with integral recycling of the actinides, because it will be necessary to remanufacture the fuel also by remote handling). Their main interest resides in the increasing of resources; interest with a very distant timescale if fast spectrum uranium systems develop normally, with closer timescales, in the opposite case. Under certain conditions, they would enable a reduction of the quantities of minor actinides produced. The thermal load of glass could be diminished, with a subsequent reduction of the needs in interim storage and disposal area. 95

94 In such a scenario, where the failure of fast spectrum systems would be postulated, thorium may only find its place in a thermal spectrum system capable of being self-sustained: the most attractive is the molten salt fuel system. The nuclear system would therefore be as follows: Accelerator Sub-critical reactor Spallation target Fission A fleet of water reactors producing plutonium; A fleet of thermal neutron molten salt reactors, started with plutonium produced in the former. Protons 1 GeV ma Neutrons Transmutation The thermal neutron molten salt reactors therefore seem to be an alternative to fast spectrum reactors from the point of view of a sustainable development of nuclear power. Consequently, this would require implementing two reprocessing processes, one by the aqueous method for water reactors, and the other one by the pyrochemical method for molten salt reactors. The thorium based systems therefore present some clear advantages and disadvantages. The result of this is that it is unlikely that they develop unless enormous amounts of fertile materials are required in the future. Accelerator-driven systems for the transmutation of waste The generation of electricity in a nuclear reactor is accompanied by the creation of heavy isotopes ( transuranium elements, heavier than uranium), some of which are radioactive over long periods of time. Among them, plutonium has an important energy potential, and France has chosen to extract it from spent fuels in order to recycle it in the fleet s reactors (MOX fuel). Other transuranium elements, mainly neptunium (Np), americium (Am), and curium (Cm) isotopes constitute a proportion of the high level and long lived waste (HLLL). As seen above, the isotopes of these minor actinides Np, Am and Cm are transmutable in a fast reactor. However, it is difficult to introduce high proportions of minor actinides in the fuel of critical reactors, for neutronic reasons linked to the low proportion of delayed neutrons and to the low Doppler effect associated with these isotopes. For the transmutation of minor actinides, another approach consists of using, reactors operating in subcritical mode driven by accelerators: the ADS ( Accelerator Driven Systems ), also known as hybrid reactors. In these systems, the neutron balance of the reactor requires an external contribution of neutrons: the sub-criticality margin thus introduced (a few percent) would enable the use of fuels loaded with a high actinide content under satisfactory conditions of safety. Therefore fleets of double strata reactors can be envisaged: for the French fleet, an assembly of a few ADSs would assure the transmutation of minor actinides produced in the main fleet of electricity generating reactors operating with U-Pu fuel. Fig Principle of the ADS (Accelerator Driven System), or hybrid reactors. The principle of the ADS is simple: accelerated protons hit a target located in the middle of the reactor s core and produce the additional neutrons needed to complement the neutron balance of a sub-critical reactor core. The production of neutrons is carried out by the spallation process. The maintaining of neutron balance in the reactor is assured by controlling the beam s intensity. Although the principle is conceptually simple, if they are constructed, the ADS will be technologically and operationally complex installations. In an industrial ADS system comprizing a ~1 GWth reactor, the associated proton accelerator must be of a very high power (beam of protons with a power able to reach a few tens of megawatts: energy of ~1 GeV, optimal for the production of neutrons, intensity of one to a few tens of ma, according to the chosen sub-criticality). Moreover, in order to avoid a too high number of power excursions, which would shorten the life of the reactor, the authorized number of undesired acceleration shutdowns is very low (a few breakdowns per year). The reliability requirement, unusual in the normal use of accelerators by physicists, is a major challenge for constructors. Only linear accelerators should be able to supply proton beams with such performances, as the intensity of cyclotrons seems limited to a few ma. In order to produce a maximum number of neutrons, the spallation target will consist of a heavy element (rich in neutrons). The proton beam will be completely stopped in this target, which will therefore dissipate the totality of the beam power. The designing of these targets, probably liquid (lead or leadbismuth), is an important technological challenge: the strength of the entrance window crossed by the proton beam and subjected to high irradiation stresses, is essential, because it constitutes a barrier between the accelerator s vacuum and the reactor; the evacuation of the heat produced by the beam and the corrosion of the target s shell by liquid metals are also important technological issues. 96 Other avenues for the distant future: thorium cycle, hybrid systems, fusion

95 p (1 GeV) n Intra-nuclear cascade α π n n Excited nucleus Evaporation d π p Oxford, in the United Kingdom). Although of lesser power and reliability than those envisaged in the industrial ADS, these installations have supplied interesting feedback. Other installations are being prepared (the SNS neutron source at Oak Ridge (United States), that of J-PARC at KEK, near Tokyo, Japan, where experiments on transmutation are planned). γ γ Fission products Fig Spallation mechanisms. n γ Spallation residue α, β, γ decay α ADS feasibility studies are now well advanced. Work has been carried out on the national level, at CEA and in a CEA-CNRS collaboration, and on the European level, under the aegis of the Technical Working Group (TWG) and within the framework of the 5 th and now, 6 th RDFP European projects. Industrialists have participed extensively in this work. The reactor of an ADS will also be highly innovative.the transfer of the beam onto a reactor core target and the dissipation of the power produced require a very different design to that of a conventional reactor, in particular as regards the safety barriers (it seems difficult to encompass the entire accelerator in a containment). The designing of fuels incorporating large proportions of actinides must call into play highly innovative concepts into play, both out of the reactor (strong γ activity and neutrons) and in the reactor (behaviour fairly unknown). Long stay times in the reactor are necessary in order to destroy a significant proportion of minor actinides and, at the same time, actinides with a higher atomic number are produced. The technological problems linked to the behaviour in reactors of such fuels, to their manufacturing, to the permanent disposal of spent fuels or to their reprocessing and possible reconditioning for re-irradiation, are highly complex and will require a great deal of R&D. Finally, the ADS safety studies will be important because they must validate an innovative design but also the new mode, of driving reactors with accelerators. No ADS has been constructed yet since the first studies in the 50s concerning the use of accelerators in order to obtain an exterior supply of neutrons in a fission reactor. However, linear proton accelerators and spallation neutron production targets have been constructed for other purposes 34 than that of the transmutation of nuclear waste (mainly LANCSE at Los Alamos, in the United States, and ISIS, near In the CEA-CNRS collaboration, zero power experiments have been carried out at Cadarache on the critical Masurca mockup (CEA-Cadarache) and a high-intensity proton injector is in the process of being constructed at CEA-Saclay. Two significant steps could be taken within the framework of the 6 th European Research Program: the completion of a demonstration experiment consisting of coupling a proton accelerator to a reactor for the first time and the production by a group of European laboratories of a fairly detailed pre-project of a significant power demonstrator. To date, no complete economic study of the ADS concept exists, in particular because the fundamental choices regarding its elements (accelerator, spallation target, reactor type) are yet to be made. However, it is certain that the cost of an ADS would be considerably higher than that of a critical reactor because, to the almost identical reactor cost, it is necessary to add those of the accelerator and the target. The future of ADS is conditioned, firstly, by a decision regarding the continuation of the studies and establishment of the partitioning-transmutation on an international level. Secondly, if the HLLL waste management mode by partitioning/transmutation is adopted, the two transmutation techniques, in critical reactor or in sub-critical reactor (ADS) will have to be dealt with, regarding the technological and economical aspects. In any case, it will not be possible to deploy ADS on an industrial scale before a few decades, as they still require a very important R&D and demonstration effort. 34. The ADS could also be a precious tool as a neutron source for technological radiation tests of various materials and, in particular, fuels. Nuclear energy of the future: what research for which objectives? 97

96 Thermonuclear fusion Potentially, fusion energy is one of the most interesting primary energy sources, because: There are no reserve problems (it consumes deuterium and lithium used to produce tritium; these elements are abundant in nature); The fusion reaction does not produce any high level and long-lived radioactive waste; Fusion does not induce greenhouse gas effects (no production of CO 2 ); A fusion reactor is intrinsically safe (immediate disappearance of plasma in the event of malfunction; no nuclear material ). But the industrial application of nuclear fusion* is still faced with major technological challenges which will require intensive R&D prior to arriving at the electricity generation installation construction stage. Several fusion reactions of light nuclei may be used in principle to produce energy. In practice, the only reaction to have a sufficiently low energy to be considered is the nuclear fusion reaction between the nuclei of two hydrogen isotopes, deuterium (D) and tritium (T): Fig Two possible magnetic configurations to confine the plasma: cylindrical and toroidal. In both cases, the D and T ions propagate along the field lines. The toroidal configuration is the basis of most installations. D + T 4 He (the α ; 3.5 MeV) + n (14.1 MeV) This reaction can only be produced if the deuterium and tritium isotopes are completely ionised, otherwise atom-atom or atom-ion collisions, much more probable than nuclear fusion, prevent the nuclei from fusing. In the case of complete ionisation and as of a few tens of kev of kinetic energy of D and T, this reaction is produced with a significant probability by tunnel effect. It can therefore be used to produce energy, if it has been possible to maintain the D and T nuclei in interaction, that is, if the plasma formed by the deuterium, tritium and electrons from the ionisation has been kept confined and sufficiently hot. Two possible methods are offered to assure the confinement: Magnetic confinement, by which the plasma s particles (charged) are maintained confined in a finite space by a suitable magnetic field configuration; Inertial confinement, which in fact involves a compressionheating of a D-T mix by pulses from laser beams or convergent pulsed particle beams; fusion is produced and lasts as long as the compression is sufficient, the process is reproduced on each pulse. Fig Combinations of coils in a Tokamak to produce the magnetic field confinement of plasma. In magnetic confinement systems, the most developed being the tokamak type, the heating of plasma (that is maintaining the kinetic energy of the D and T nuclei at a value high enough for fusion to occur) takes place in many ways: transfer of the energy from α particles from the fusion reaction to the D+T plasma; ohmic heating induced by the plasma s 98 Other avenues for the distant future: thorium cycle, hybrid systems, fusion

97 electric current; heating by high frequency electromagnetic waves or heating by neutral particle injection. The state-of-the-art on fusion by magnetic confinement So that fusion can be used as an energy source, it is necessary that the energy produced by the fusion be greater than that injected to heat and maintain the plasma (Q = Pfus/Pext >1). This system, called breakeven is expressed by a constraint of the type: n. τ E > f (T) Lawson parameter; n. τe (10 20 m -3 s) Ignition ITER where n is the density of plasma, τ E the confinement time and T its temperature. Considerable progress has been made in the last few decades on the fulfillment of this criterion. By way of example, the Tore- Supra installation in Cadarache now produces plasmas confined over several minutes, and the European JET installation, in Culham, England, is close to meeting this Q>1 constraint. Many technological problems, however, are yet to be solved prior to envisaging the construction of an industrial installation, in which the α particles produced by the reaction would be sufficient to heat the plasma. These technological problems are of three orders: Central ion temperature (kev) Fig Performances reached by existing or project installations, in the Lawson vs. temperature parameter diagram. It will be noted that ITER* is located at the limit of the ignition zone (a α particles from the fusion will assure 2/3 of the plasma s heating). The toughness of the materials in contact with the plasma: In an industrial system, the first lining must evacuate a very high power density (it may locally exceed 20 MW/m 2 ), support the very high neutron fluxes which will cross it in order to go into the tritigenous blankets and enable the evacuation of a large quantity of gaseous helium originating from the fusion reaction. The development, in the 80s, of the divertor, a special magnetic configuration making it possible to better manage the fluxes to be evacuated from the plasma, has contributed a great deal to the progress accomplished during the last two decades. The production of tritium in the lithium tritigenous blankets: In such systems, the neutrons, originating from the fusion reaction continuously produce the tritium consumed in the plasma, by interaction with the lithium blankets ( tritigenous blankets ), thus avoiding the storage and handling of this radioactive element. The overall quantity of tritium present in an industrial type installation is therefore only a few grams. However, tritium diffuses easily and the control of its diffusion is an important safety aspect. The minimisation of the activation of the blanket materials: The neutrons interacting with the walls cause the appearance of radioactive elements, via nuclear reactions. The choice of the composition of the blanket materials must be such that it minimises, in level and in time, the production of the reactivity induced. Inertial confinement-heating only seems possible with photon beams (lasers) or heavy ions, but, even in this case, the scientific and technical problems to be solved do not enable an industrial-scale implementation in a foreseeable future. The ITER project and nuclear fusion outlook The purpose of the global ITER nuclear fusion experimental reactor project is to scientifically and technically demonstrate that it is possible to use fusion to produce energy. The partners are the European Union, Russia, Japan, the United States, China and South Korea. The installation (Fig. 113) will be of the tokamak type.with size and performances similar to the industrial reactors envisaged (Q = 10, heating of the plasma to 66% by α particles), it will Nuclear energy of the future: what research for which objectives? 99

98 enable the research which is still necessary on materials and the operation of a fusion reactor to be carried out in a realistic configuration. The Cadarache site was retained to host this 4.5 bn installation. The construction time is 12 years and ITER should be operated during approximately twenty years. If the results gathered and the studies of materials confirm the scientific and technological possibility of using nuclear fusion for the production of energy, an industrial production reactor prototype, studied in parallel to the ITER operation, could then be constructed. Even then, the road leading to the industrial exploitation of nuclear fusion will be long, because the economic competitiveness of this mode of energy production still remains to be demonstrated. Central solenoid Blanket element Toroidal field coils Vacuum chamber Poloidal field coils Cryostat Heating of the plasma Divertor Support columns Cryopump Fig Diagram of the ITER installation, global scientific and technical validation project for the use of fusion for the production of energy. 100 Other avenues for the distant future: thorium cycle, hybrid systems, fusion