[en] The potential sources of tritium contamination of the helium-coolant of ceramic breeder blankets have been evaluated in a previous paper for the specific case of the European BIT DEMO blanket. This evaluation associated with a rough assessment of the permeability to tritium of the tubing of helium-heated steam generators confirmed that the control of tritium losses to the steam circuit is a critical issue for this class of blanket requiring developments in three areas: (1) permeation barriers, (2) tritium recovery processes maintaining a very low concentration in tritiated species in the coolant, and (3) methods for controlling the chemistry of the coolant. Consequently, in order to define the specifications of these developments, a detailed evaluation of the permeability to tritium of helium-heated steam generators (SGs) was performed, which will be reported in this paper. This study includes the definition of the thermal-hydraulic operating conditions of the SGs through thermodynamic cycle calculations, and its thermal-hydraulic design. The obtained geometry, area and temperature profiles along the tubes are then used to estimate, based on relevant permeability data, the tritium permeation through the SG as a function of the composition in tritiated species of the coolant. The implications of these results, in terms of requirements for the considered tritium control methods, will also be discussed on the basis of expected limits in tritium release to the steam circuit

[en] Within the framework of the European test-blanket program, CEA and ENEA are jointly developing a DEMO-relevant, helium-cooled, Breeder-Inside-Tube (BIT) ceramic blanket. Two ceramics are possible breeder material candidate: LiAlO₂ and Li₂ZrO₃. Despite the design has been originally developed for aluminate, the CEA has recently focused its work on zirconate. This concept blanket segments are made by a directly-cooled vacuum-tight steel box which contains banana-shaped poloidal breeder modules arranged in rows (6 rows in an outboard segment and 4 rows in an inboard one). A breeder module consists of a pressure vessel containing a bundle of breeder rods surrounded by baffles. Each one of the rods is made-up of a steel tube containing a stack of annular pellets of sintered lithium-zirconate, through which flows helium (the tritium purge gas). The thermo-mechanical analysis has shown that the thermal gradient in the ceramics can be kept at acceptable levels despite the poorer out-of-pile thermo-mechanical properties of zirconate compared to aluminate. Moreover, the neutronic analysis has shown that, besides the maintained tritium-breeding self-sufficiency capability of this blanket, the lower lithium burn-up could be an indication that the zirconate characteristics remains more stable after long term irradiation (i.e., close to the end-of-life fluence of 5 MWa/m²)

[en] The Helium Cooled Pebble Bed (HCPB) blanket is one of the two European breeding blanket concepts for DEMO. In the framework of selecting reference blankets for the European Power Plant Conceptual Study (PPCS) performance limits of the HCPB concept have been explored. This paper shows the improved, moderate-technological-extrapolation HCPB blanket design including recent changes to the breeding zone that counteract a recuperator effect in the cooling plates. It highlights the attractive features of the design, like single-size Beryllium beds, large admissible neutron wall loads (NWL) and much improved electrical efficiency. In addition to two reference blanket cases proposed in the study, a third blanket based on the small NWL in the PPCS reactor is put forward. Finally, steam cycle efficiencies calculated with an up-to-date power plant code correct previous values for the improved HCPB blanket upwards by about 2%

[en] One of the missions of the International Thermonuclear Experimental Reactor (ITER) project is to serve as a test bed for various test blanket concepts. Part of the design for these test modules is to assess their impact on the safety of ITER. This paper presents a set of safety analyses, which were performed on a consistent basis of analysis assumptions and safety requirements. The four basic concepts investigated are a helium-cooled ceramic breeder, a water-cooled ceramic breeder, a water-cooled Pb-17Li breeder, and a liquid lithium self-cooled concept. The analysis performed so far has confirmed the expectation of little safety impact from the test modules, except in a few areas. The use of liquid lithium poses well-known risks of chemical reactivity, which need to be further addressed. For the water-cooled Pb-17Li concept the need was confirmed for further R and D on reactivity of Pb-17Li with water to quantitatively assess bounding scenarios. For the water-cooled ceramic test module the analysis has identified the issue of excessive Be-steam reactions, which needs further investigations

[en] The European Safety Analysis Platform (ESAP) is a computational platform, with the objective to perform an integrated core and safety analysis of nuclear reactor systems. The current components of ESAP include: MCNP5 for neutron transport calculations, COBRA for core and sub-channel analysis and FRETA for fuel thermal-mechanical behaviour studies. Currently, the platform is able to perform safety design studies and has been applied to Sodium cooled Fast Reactors for both fuel assemblies and for the whole core. This will be extended later to other reactor types including the Lead-cooled Fast Reactor. The objective is to be able in the long run to perform safety design and transient analysis of most nuclear energy systems. The paper describes the present structure of the platform as well as its architecture and data exchange. Examples of first applications of ESAP within the European Sodium Fast Reactor (ESFR) project are also provided. (authors)

[en] In this overview paper, the following questions are addressed: (1) What are the challenges facing the European Union nuclear fission research community in the short (today), medium (2010) and long term (2040)? (2) What kind of research and technological development (RTD) does Euratom offer to respond to these challenges, in particular in the area of reactor systems and fuel cycles? In the general debate about energy supply technologies there are challenges of both a scientific and technological (S/T) as well as an economic and political (E/P) nature. Though the Community research programme acts mainly on the former, there is nevertheless important links with Community policy. These not only exist in the specific area of nuclear policy. It is shown in the particular area of nuclear fission, to what extent Euratom research, education and innovation ('Knowledge Triangle: Education, Research, and Innovation') respond to the S/T challenges: (1) sustainability, (2) economics, (3) safety, and (4) proliferation resistance. At the European Commission (EC), the research related to nuclear reactor systems and fuel cycles is principally under the responsibility of the 2 Directorates Generals (DG) DG Research (RTD, located in Brussels), which implements and manages the programme of 'indirect actions', and the DG Joint Research Centre (JRC, headquarters in Brussels and 7 scientific institutes in 5 Member States) which carries out 'direct actions' in their own laboratories. In this HTR-2006 introductory paper, the emphasis is on the indirect and direct actions of the 6^th Euratom research framework programme 2003-2006, FP-6, with special emphasis on V/HTR Generation IV research. (orig.)

[en] The European Strategic Energy Technology (SET) Plan recognizes the need to develop new energy technologies, in order to reduce greenhouse gas emissions and secure energy supply in Europe. Besides renewable energy and improved energy efficiency, a new generation of nuclear power plants and innovative nuclear power applications can play a significant role to achieve this goal. The JRC Institute for Energy 'Safety of Future Nuclear Reactors' (SFNR) Unit is engaged in experimental research, numerical simulation and modelling, scientific, feasibility and engineering studies on innovative nuclear reactor systems. This also represents a significant EURATOM contribution to the Generation IV International Forum. Its activities deal with, among others, the performance assessment of innovative fuels and materials, development of new reactor core concepts and safety solutions, and knowledge management and preservation. Special attention is given to fast reactor concepts, namely the sodium (SFR) and lead (LFR) cooled reactors. Recognizing the maturity of the SFR technology, the European Sustainable Nuclear Energy Technology Platform (SNETP) considers a prototype SFR to be built as a next-step towards the deployment of a first-of-a-kind Gen IV SFR. This paper gives an overview of current research preformed at JRC-IE with emphasis on the work performed in the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) within the European Commission's Seventh Framework Program. (authors)

[en] The HFR Petten is a key research reactor in Europe and in the course of the last decades several dedicated irradiation facilities have been developed and successfully operated. For example, as regards materials irradiation, in the frame of the European Network AMES (Ageing Materials and Evaluation Studies), the irradiation behaviour of reactor pressure vessel (RPV) steels, and thermal annealing efficiency and sensibility to re-irradiation damage are being investigated. Today the rig is a very promising tool for GIF research on materials. Similarly the QUATTRO rig has been tailored to GIF research. Dedicated irradiation rigs are also designed and operated by the IE to test advanced fuels; e.g. HTR fuel. An overview of the developed facilities is given in the following paper as well as examples of achieved results. (author)

[en] In the european fusion programme of 1999 preparatory work (preparation of a power plant conceptual study - availability, PPA) has been carried out for a fusion power plant study that is planned to start in 2000. This study will focus on the commercial attractiveness of a fusion plant, particularly achievable power level, net efficiency and availability. Part of the activity at the Forschungszentrum Karlsruhe has been the further development of the helium cooled pebble bed (HCPB) blanket for DEMO as ''improved HCPB'' (subtask PPA 2.3). The modified concept allows for the height of breeder pebble beds to be reduced and thus for larger power densities to be accommodated. Also, mono-disperse beryllium pebble beds can be used. The net electric efficiency of the blanket was raised by almost 7 points to about 37% due to increased coolant temperature gain, reduction of pressure losses in the blanket and enhanced energy conversion in the proposed steam process. The good availability of the DEMO-HCPB that was shown in earlier studies is expected to carry over to the I-HCPB. (orig.)

[en] The water-cooled Pb-17Li blankets represent one of the blanket lines selected within the European Union for DEMO-relevant design and R ampersand D activities. This paper gives a presentation of the reference conceptual design for water-cooled Pb-17Li DEMO blankets and an overview on the results of its performance assessments. Moreover, a critical discussion about the technical aspects requiring further improvements and/or modifications is performed taking into account the present status of the associated R ampersand D. This concept appears to be a very promising candidate for a DEMO reactor breeding blanket