[en] British Nuclear Fuels Limited's (BNFL's) Waste Vitrification Plant at Sellafield is responsible for processing high-level radioactive waste. This history of the process development and the operation of the plant are outlined. Early operation experience revealed that the severe in-cell operating environment was causing high unreliability in several key items of equipment. Subsequent improvements to overcome these problems are outlined, leading to increased plant throughput overall. (UK)

[en] The reprocessing of irradiated nuclear fuel at British Nuclear Fuels (BNFL) Sellafield site consists of a number of relatively self-contained activities carried out in separate plants across the site. The physical conditions and time scales applied in reprocessing and storage make it relatively benign. The potential for minor releases of radioactivity under fault conditioning is minimised by plant design definition of control procedures, training and supervision. The risks to both the general public and workforce are shown to be low with all the safety criteria being met. Normal operating conditions also have the potential for some occupational radiation exposure and the plant and workers are monitored continuously. Exposure levels have been reduced steadily and will continue to fall with plant improvements. (U.K.)

[en] In 2011, ETSON published the ''Position Paper of the Technical Safety Organizations: Research Needs in Nuclear Safety for Gen 2 and Gen 3 NPPs''. This paper, published only a few months after the Fukushima-Daiichi severe accidents, presented the priorities for R and D on the main pending safety issues. It was produced by the ETSON Research Group (ERG) that has the mandate of identifying and prioritizing safety research needs, sharing information on research projects in which ETSON members are involved, defining and launching new research projects and disseminating knowledge among ETSON members. Six years after this publication, many R and D international projects finished in diverse frames, and other ones have started. In particular a lot of work was done (and is going on..) on the analysis of the Fukushima-Daiichi severe accidents. Meanwhile a roadmap on research on Gen. 2 and 3 nuclear power plants (NPP), including safety aspects, was produced by the NUGENIA association, followed by a more detailed document as ''NUGENIA global vision''. It was also demonstrated that the ETSON R and D priorities were consistent with the implementation of the 2014 Euratom Directive on safety of nuclear installations.

[en] It is a major advantage to those that must make decisions about the implementation of an emergency plan following an accident at a nuclear power plant if the likely release of activity is predicted before it occurs. To this end, a software module, which provides a rapid estimate of the source term to the environment, has been developed by NNC Ltd. Termed the RODOS STM, the software has been developed such that it can be used with the RODOS system. Operating outside of RODOS, the software remains a useful source term prediction tool. The software employs Bayesian analysis techniques, and the results of level 1 and level 2 probabilistic safety analysis, to calculate the probability of the possible releases of activity into the environment and the potential magnitude of those releases, given a set of observations about the status of the NPP. The software has been developed using the Sizewell 'B' PWR design. (author)

[en] In the frame of the RAPHAEL European Project, the back end of the fuel cycle is addressed concerning mainly open cycles : direct disposal of spent fuel or of separated constituents (fuel matrix, particles, compacts or pebbles). One of the objectives is to obtain more precise experimental data allowing to predict the fuel behaviour for direct geological disposal conditions. The global effort is shared between both bibliographic and experimental work including : - the establishment of an international comparison of contamination characterization levels and disposal specifications to propose a list of conditioning and/or decontamination requirements, - a compilation of existing data about HTR spent fuel characterization, - the acquisition of physical, chemical, radiological states and inventory data of irradiated and fresh fuels with LEU kernels. The characterization programme will be carried out in the Atalante Facility (CEA) in France, using samples provided by Juelich center in Germany (FZJ). Both destructive and non destructive analysis will be performed with pebbles and coated particles. This paper describes in detail both characterization techniques and objectives and gives an overview about the equipments that will be used. (author)

[en] In the frame of the RAPHAEL European Project, the back end of the fuel cycle is addressed concerning mainly open cycles : direct disposal of spent fuel or of separated constituents (fuel matrix, particles, compacts or pebbles). One of the objectives is to obtain more precise experimental data allowing to predict the fuel behaviour for direct geological disposal conditions. The global effort is shared between both bibliographic and experimental work including : - the establishment of an international comparison of contamination characterization levels and disposal specifications to propose a list of conditioning and/or decontamination requirements, - a compilation of existing data about HTR spent fuel characterization, - the acquisition of physical, chemical, radiological states and inventory data of irradiated and fresh fuels with LEU kernels. The characterization programme will be carried out in the Atalante Facility (CEA) in France, using samples provided by Juelich center in Germany (FZJ). Both destructive and non destructive analysis will be performed with pebbles and coated particles. This paper describes in detail both characterization techniques and objectives and gives an overview about the equipments that will be used. (author)

[en] For various countries, the direct disposal of high level nuclear fuel wastes is a key option for the back-end of the fuel cycle. For HTR/VHTR reactors this is assumed for the introductory phase of this reactor system. However, closed fuel cycles or a separation of spent coated-particles from the graphite moderator and specific treatment, conditioning and disposal of these waste streams are also possible. In the European Community project 'RAPHAEL', fuel waste performance is going to be studied in depth, including post-irradiation fuel characterization, analysis of the stability and failure mechanism of coatings and of fuel kernels and overall performance of waste packages with compact fuel and/or only with fuel particles in geological disposal environments. Different confinement matrices for separated fuel particles (vitrification, SiC, ZrO₂) have been adapted to limit release of radionuclides into groundwater at low temperatures over geological time spans. The investigations are limited to Low-Enriched Uranium (LEU) fuel with uranium oxide and uranium oxycarbide kernels that will allow higher burn-up, but may be more susceptible to leaching. (authors)

[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)