[en] During last few years, the interest in the innovative, Liquid Salt cooled - Very High Temperature Reactor (LS-VHTR), has been growing. The preconceptual design of the LS-VHTR was suggested in Oak Ridge National Laboratory (ORNL) [1] and nowadays, several research institutions contribute to the development of this concept. The LS-VHTR design utilises a prismatic, High Temperature Reactor (HTR) fuel [2] in combination with liquid salt as a coolant. This connection of high-performance fuel and a coolant with enhanced heat transfer abilities enables efficient and economical operation. Main objective of the LS-VHTR operation may be either an efficient electricity production or a heat supply for a production of hydrogen or, combination of both. The LS-VHTR is moderated by graphite. The graphite matrix of the fuel blocks, as well as the inner and outer core reflectors serve as a thermal buffer in case of an accident, and they provide a strong thermal feedback during normal reactor operation. The high inherent safety of the LS-VHTR meets the strict requirements on future reactor systems, as defined by the Gen IV project. This work, purpose, scope, contribution to the state-of-art: The design, used in the present work is based on the first ORNL suggestion [1]. Recent study is focused on comparison of the neutronic performance of two types of fuel in the LS-VHTR core, whereas, in all previous works, only uranium fuel has been investigated. The first type of fuel, which has been employed in the present analysis, is based on the spent Light Water Reactor (LWR) fuel, whereas the second one consists of enriched uranium oxide. The results of such a comparison bring a valuable knowledge about limits and possibilities of the LS-VHTR concept, when employed as a spent fuel burner. Method:It is used a 3-D drawing of the LS-VHTR core, which contains 324x10 hexagonal fuel blocks. Each fuel block contains 216x10 fuel pins, which consists of TRISO particles incorporated into a graphite matrix. The external radius of a TRISO particle has been set to 410μm, the radius of the fuel kernel to 150μm, in case of plutonium fueled core, and 215μm in case of uranium fueled core. The core was modelled in stochastic, three-dimensional code MCNP, version 4c3, in the finest detail. First, an under moderated core setup was found for both types of fuel by modifying the fuel to moderator ratio; then, the void and the thermal coefficients of reactivity were investigated. Few single - component molten salts were involved in the study of the void effect, in order to estimate worth of these components; NaF, BeF₂, LiF, ZrF₄. As a reference multi-component salt, Li₂Be₄F, referred to as FLiBe, was investigated. Results: It can be seen that removing BeF₂ from the core brings a negative reactivity contribution, while other three components, NaF, LiF and ZrF₄ would in a mixture contribute to the reactivity positively. Voiding FLiBe, which is a mixture of 66% of LiF and 34% BeF₂, is equivalent to a negative reactivity insertion. Both the moderator and the fuel temperature coefficients of reactivity are large and negative for both plutonium and uranium fueled core. In the operational temperature interval (1200 K for graphite and 1500 K for fuel), the total temperature feedback is - 7.82 pcm/K for the plutonium fueled core and -2.47 pcm/K for the uranium fueled core. This results show, that the LS-VHTR core has a potential to meet the basic safety requirements as both uranium, and spent LWR fuel burner. References: [1] D. T. Ingersoll, L. J. Ott, J. P. Renier, S. J. Ball, W. R. Corwin, C. W. Forsberg, D. F. Williams, D. F. Wilson, L. Reid, G. D. Del Cul, P. F. Peterson, H. Zhao, P. S. Pickard, E. J. Parma. Status of Preconceptual Design of the Advanced High-Temperature Reactor. (ORNL, The United States of America, Tennessee 2004). [2] A. Talamo, W. Gudowski, F. Venneri, Annals of Nuclear Energy, 31, 173-196 (2004), The burnup capabilities of the Deep Burn Modular Helium Reactor analyzed by the Monte Carlo Continuous Energy Code MCB

[en] The following contributions are covered: The Czech power system and its future; Nuclear power and local development; Preconditions for the completion of the Temelin NPP; Nuclear opportunity for the Czech Republic; Development trends in nuclear power; Safe operation of nuclear power plants; Safe radioactive waste disposal in the Czech Republic; and Renaissance of nuclear power. (P.A.)

[en] The PUMA project -the acronym stands for 'Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors'- was a Specific Targeted Research Project (STREP) within the EURATOM 6th Framework Program (EU FP6). The PUMA project ran from September 1, 2006, until August 31, 2009, and was executed by a consortium of 14 European partner organisations and one from the USA. This report serves 2 purposes. It is both the 'Publishable Final Activity Report' and the 'Final (Summary) Report', describing, per Work Package, the specific objectives, research activities, main conclusions, recommendations and supporting documents. PUMA's main objective was to investigate the possibilities for the utilisation and transmutation of plutonium and especially minor actinides in contemporary and future (high temperature) gas-cooled reactor designs, which are promising tools for improving the sustainability of the nuclear fuel cycle. This contributes to the reduction of Pu and MA stockpiles, and also to the development of safe and sustainable reactors for CO₂-free energy generation. The PUMA project has assessed the impact of the introduction of Pu/MA-burning HTRs at three levels: fuel and fuel performance (modelling), reactor (transmutation performance and safety) and reactor/fuel cycle facility park. Earlier projects already indicated favourable characteristics of HTRs with respect to Pu burning. So, core physics of Pu/MA fuel cycles for HTRs has been investigated to study the CP fuel and reactor characteristics and to assure nuclear stability of a Pu/MA HTR core, under both normal and abnormal operating conditions. The starting point of this investigation comprised the two main contemporary HTR designs, viz. the pebble-bed type HTR, represented by the South-African PBMR, and hexagonal block type HTR, represented by the GT-MHR. The results (once again) demonstrate the flexibility of the contemporary (and near future) HTR designs and their ability to accept a variety of Pu- and Pu/MA-based fuels (possibly in combination with thorium), and to obtain a significant reduction of the Pu- respectively Pu/MA content, while maintaining, to a large extent, the well-known standard (U-fuelled) HTR safety characteristics. However, this will require some changes in the reactor design. Studies have furthermore shown that fuel with a 'diluted' kernel ('inert-matrix') improves the transmutation performance of the reactor. A study on proliferation resistance, taking into account several possible proliferation pathways, highlights that a prismatic (V)HTR core would be amenable to conventional safeguards accounting and verification procedures, with fuel blocks accounted for individually in the same way as LWR fuel assemblies. However, a modified approach would be needed in pebble bed cores because of the impracticability of accounting for individual fuel spheres. When dealing with minor actinide bearing fuel helium generation is an important issue. Experiments have shown that He will be released from the kernel, but not from fresh kernels. Indeed, fresh fuel has shown a remarkable stability up to 2500 degrees C. For modelling purposes, 100% release of helium from the kernel is justified. The diluted kernel concept was first invoked by Belgonucleaire brings many benefits. The fuel modelling studies have clearly indicated the advantages that can be gained by dilution. Essentially, for a given buffer layer thickness, more volume is available to accommodate the CO and He released. Chemical thermodynamic models have been deployed to design a kernel that will show limited CO production. The most important point here is that substoichiometric Pu or Am oxides are essential. Further improvement can be achieved by chemical buffering of the fuel by the addition of Ce sesquioxide, which takes up oxygen to form the dioxide. Ultimately any coated particle design must be validated in an irradiation test. Though not possible to perform an irradiation programme in the PUMA project, the feasibility of such a programme has been demonstrated, and the initial data needed to launch such a test has been generated. Pu/MA transmuters are envisaged to operate in a global system of various reactor systems and fuel cycle facilities. Fuel cycle studies have been performed to study the symbiosis to other reactor types/systems, and to quantify waste streams and radio toxic inventories. This includes studies of symbiosis of HTR, Light Water Reactor (LWR) and Fast Reactor (FR) systems, as well as the assessment of the technical, economic, environmental and socio-political impact. It is e.g. shown that a Pu/MA-loaded HTR may have a considerable, positive impact on the reduction of the amount of TRU in disposed spent fuel and high level waste.

[en] The PUMA project, a Specific Targeted Research Project (STREP) of the European Union EURATOM 6. Framework Program, is mainly aimed at providing additional key elements for the utilisation and transmutation of plutonium and minor actinides (neptunium and americium) in contemporary and future (high temperature) gas-cooled reactor design, which are promising tools for improving the sustainability of the nuclear fuel cycle. PUMA would also contribute to the reduction of Pu and MA stockpiles and to the development of safe and sustainable reactors for CO₂-free energy generation. The project runs from September 1, 2006 until August 31, 2009. PUMA also contributes to technological goals of the Generation IV International Forum. It contributes to developing and maintaining the competence in reactor technology in the EU and addresses European stakeholders on key issues for the future of nuclear energy in the EU. An overview is presented of the status of the project at mid-term. (authors)