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[en] The current open nuclear fuel cycle uses only a few percent of the energy contained in uranium. This efficiency can be greatly improved through the recycling of spent fuel (as done today in France for instance), including, in the longer term, multi-recycling strategies to be deployed in fast reactors. In this context, and in the continuity of the FP7 EURATOM SACSESS project, GENIORS addresses research and innovation in fuel cycle chemistry and physics for the optimization of fuel design in line with the strategic research and innovation agenda and deployment strategy of SNETP, notably of its ESNII component. GENIORS focuses on reprocessing and fuel manufacture of MOX fuel potentially containing minor actinides, which would be reference fuel for the ASTRID and ALFREDO demonstrators. More specifically, GENIORS carries out research and innovation for developing compatible techniques for dissolution, reprocessing and manufacturing of innovative oxide fuels, potentially containing minor actinides, in a “fuel to fuel” approach taking into account safety issues under normal and mal-operation. The different promising options developed in SACSESS are currently further developed to address the specific challenges of GEN IV. For delivering a full picture of a MOX fuel cycle, GENIORS works in close collaboration with the INSPYRE project on oxide fuels performance. By implementing a three step approach (reinforcement of the scientific knowledge => process development and testing => system studies, safety and integration), GENIORS will lead to the provision of more science-based strategies for nuclear fuel management in the EU. It will allow nuclear energy to contribute significantly to EU energy independence. This paper presents the strategy and current results of GENIORS. (author)
[en] The Th fuel cycle is attracting interest again globally because of its advantages over the current Pu fuel cycle, such as breeding fissile 233U from fertile 232Th without using a fast reactor, lower minor actinide production and higher Pu burning. However, there are some concerns, such as the small critical mass of the bred 233U. Using 234U, which is not considered an important isotope, may overcome some problems with the Th fuel cycle. In this study, the effect and roles of 234U in the Th fuel cycle were surveyed from the perspectives of proliferation resistance (PR), fuel burn-up, and breeding in single and multiple cycles. Increasing the 234U isotope ratio increases bare critical mass, which in turn increases PR by increasing the heat generation and radiation dose rate from 232U and their daughter nuclei. The effects of the moderator-to-fuel ratio, neutron energy spectrum, and neutron flux (linear power density) on criticality were estimated. 234U was fissile in the faster neutron energy spectrum, which can increase the fuel burn-up under some conditions. A higher fuel burn-up is preferable to increase the 234U isotopic ratio. For multiple cycles, the breeding ability of 234U was higher with a softer neutron energy spectrum (33.3% at the end of the fifth cycle), but the mass balance was worse. When 234U was used with a harder neutron energy spectrum, the 234U isotopic ratio was as high as 23.6%, but the mass balance was better. The role of 234U in Th has not been thoroughly investigated until now, but this study has revealed the importance of 234U, which may lead to the development of a new Th fuel cycle. (author)
[en] To develop nuclear energy is inevitable choice for China to meet the requirement of decreasing greenhouse gas emission, at the same time of economic and society development. To ensure sustainable development of nuclear energy, closed nuclear fuel cycle strategy based on fast reactor has to be adopted. Both of recent and next R&D activities of nuclear fuel cycle back-end were introduced in the paper, such as: — Nuclear energy development and spent fuel accumulation, including fast reactor and ADS development aiming at transmutation long-lived nuclides; — Commissioning of Reprocessing Pilot Plant for PWR spent fuel, development of advanced PUREX process and hot test of separation both U and Pu in CRARL (China reprocessing and radiochemistry laboratory); — Minor Actinides separation on laboratory scale; — Investigation on vitrification of high-level liquid waste, high level waste disposal and its programme. (author)
[en] India has adopted a „closed fuel cycle‟ considering spent fuel a material of resource. This has enabled not only optimally utilising the scarce resource of Uranium but also helped in efficient management of radioactive-waste and opening the possibilities for tapping the energy of various useful radio-isotopes present in waste for societal benefits which otherwise are not available in nature. Reprocessing of spent fuel enables in recovering of fuel and recycling them to future reactor for utilising as fuel. Such recovery and recycling of fuel material in reactor to generate power not only helps in ensuring the energy security of the country but also helps in reducing the rad-waste volume meant for geological disposal to a great extent. Spent fuel reprocessing results in recovery of more than 95% of material and hence generates a very small amount of high level liquid waste (HLLW) which is significantly lower than the direct disposal of spent fuel in case of „open fuel cycle‟.The HLLW, characterised by high concentration of radioactivity in combination with presence of long lived minor actinides, poses the challenge for its safe management. The HLLW is vitrified in suitable glass matrix and interim stored for removal of decay heat. The advantage of vitrification of HLLW into vitreous matrix is to immobilise the radioactivity in chemically durable form ascertaining containment and isolation of radioactivity from the human environment for extended period of time. HLLW contains many valuable radio-nuclides such as Cs-137, Sr-90, Ru-106 etc which have various societal applications in the field of industry and healthcare. India has put a step forward in implementation of advance fuel cycle by recovering the valuable radionuclides from HLW and deploying them for various societal applications. Separation science has played a key role in selective recovery of these radio-nuclides in pure form from HLLW. Recovery of Cs from HLLW using solvent extraction based system enabled use of Cs in non-dispersible glass form for irradiation purpose. Recovery of Sr-90 from HLLW was also demonstrated to milk out the radio-pharmaceutical grade Y-90 for radiopharmaceutical applications. Recently, Ru-106 has been recovered from HLLW to produce Ru plaque for eye cancer treatment. Reprocessing of spent fuel for recovery of heavy metals followed by extraction of useful radioisotopes reduces the waste volume immensely prior to isolation and their eventual disposal. The paper outline, the practices being adopted in India for management of high level radioactive waste. A brief description covering the important aspects of waste management like the sources of HLW, composition details along with management strategy is given below. (author)
[en] Mindful of possible future limitations on the availability of uranium, the introduction of the thorium fuel cycle is potentially a complementary source of nuclear energy. This publication assimilates current knowledge of thorium geology and mineralization into a brief account on the worldwide occurrence of thorium resources. Although thorium is currently not commercially viable as a fuel, it is important to pre-emptively assess thorium related information should that situation change. Thus, the publication provides an overview of the variety of natural thorium deposit types with associated thorium geology and thorium resources. It reviews available data on thorium occurrences/deposits and thorium resources and presents a classification of deposits according to geological and economic criteria.
[en] The current growth of the energy demand, the perspective of a pronounced increment for the next decades, added to the near depletion of the fossil fuels has made necessary finding a sustainable energy supply. Nuclear energy is presented as an important energy source to meet global energy needs in the near future, without adverse impacts to the environment. However, it faces substantial challenges to be successful as sustainable energy source. Pebble Bed Very High Temperature advanced systems together with fuel cycles based in Thorium present significant perspectives to assume the future nuclear energy. In this paper the main advantages of the use a Generation IV Very High Temperature Hybrid System (VHTHS) using a variety of fuel cycles based on Thorium (232Th-233U, 232Th-239Pu and 232Th-U) under a deep burn scheme are studied. The conceptual design of the VHTHS composed of a Very High Temperature Pebble Bed Reactor (VHTR) and two Pebble Bed Accelerator Driven Systems (ADSs) is analyzed. Parameters such as the isotopic composition, the Minor Actinides stockpile, the nuclear fuel breeding, the percent fissile fuel and the radiotoxicity of the long-lived wastes are determined in order to know the influence of using Thorium based fuel cycles and ADSs as a second stage in the VHTHS. The MCNPX version 2.6e was used for the neutronic calculations. (author)
[en] The Nuclear Fuel Cycle Simulation System (NFCSS) is a scenario based computer simulation tool that can model various nuclear fuel cycle options in various types of nuclear reactors. It is very efficient and accurate in answering questions such as: the nuclear mineral resources and technical infrastructure needed for the front end of the nuclear fuel cycle; the amounts of used fuel, actinide nuclides and high level waste generated for a given reactor fleet size; and the impact of introducing recycling of used fuel on mineral resource savings and waste minimization. Since the first publication on the NFCSS as IAEA-TECDOC-1535 in 2007, there have been significant improvements in the implementation of the NFCSS, including a new extension to thorium fuel cycles, methods to calculate decay heat and radiotoxicity, and demonstration applications to innovative reactors.
[en] This section presents the case study from the Russian Federation on modelling of multilateral NES using the IAEA’s model MESSAGE. The configuration of regional NES simulated in the study was jointly designed in the INPRO collaborative project SYNERGIES by participants from Armenia, the Russian Federation and Ukraine. Belarus did not directly participate in the SYNERGIES project, although experts from Belarus, being active participants of the INPRO project, have provided the data necessary for the case study. To simulate regional configuration and incorporate a complete spectrum of NFC elements, the extended capabilities of the MESSAGE code presented elsewhere were used. Regarding regional NES cooperation scenarios, the national partners involved in modelling the case study for regional NES represent different options of nuclear power development and deployment. In accordance with the terminology introduced in the INPRO collaborative project GAINS, they could be assigned to different nuclear energy strategy groups. The Russian Federation belongs to the nuclear energy strategy group, which pursues a general strategy of spent nuclear fuel recycling. This group plans to build, operate and manage used fuel recycling facilities and permanent geological disposal facilities for highly radioactive waste. Armenia and Ukraine belong to a nuclear energy group, which either follows a strategy of direct disposal of the SNF or that of its reprocessing abroad. This group plans to build, operate and manage permanent geological disposal facilities for highly radioactive waste (in the form of used fuel and/or reprocessed waste) and/or to work in collaboration with another group of countries to have its fuel recycled. Belarus belongs to a group, which has a general strategy of sending spent fuel abroad for recycling or disposal, although the ultimate back-end strategy is still undecided. The scenarios developed in the SYNERGIES project can serve as a hypothetical example of a potentially ‘win–win’ cooperation in the NFC front and back ends.
[en] China is striving to achieve sustainable environmental development for the future. Developing nuclear power is one of the important options for China’s energy supply structure optimization. The Government of China has consistently advocated the development of nuclear power on the basis of nuclear safety. The Qinshan Nuclear Power Plant was the first nuclear power plant established in China and was put into operation in 1991. Since then 28 units had been put into commercial operation by the end of 2015, bringing the total installed nuclear capacity to 26.42 GW(e). Twenty four of these units are based on PWR technology, corresponding to 24.98 GW(e) in terms of power production. China has a far-reaching nuclear power development vision. In the next five years, about 30 GW(e) of nuclear capacity will be put into operation and more than 30 GW(e) capacity will be under construction. The nuclear power capacity is estimated to be 58 GW(e) by 2020. The top level scenario of China's nuclear power development is a three-step strategy of ‘thermal reactor–fast reactor–fusion reactor’ which was adopted in the 1990s. China has long been focusing on the development of nuclear power technology. The Generation III nuclear power technologies, for example HPR1000 and CAP1400, have been developed domestically. Significant R&D efforts have also been put into developing the technology of Generation IV nuclear energy systems, such as fast reactor with sodium coolant (SFR), molten salt reactor (MSR), etc. China will become a technology exporting country and will contribute to the development of global nuclear energy in the future. In addition, the role of nuclear energy is increasingly dependent on the economic environment of energy markets and the competitiveness of the alternative energy technologies. As such, the entire energy/electricity market should be modelled and the role of nuclear energy in the entire system setting should be assessed. However, using MESSAGE for the entire electricity market in China would be very challenging owing to the size and complexity of the market.
[en] The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) at IAEA has carried out assessment studies on global, regional and country levels based on INRPO methodology for developing a global vision of nuclear energy sustainability for the 21st century. Particularly, the GAINS project developed a global architecture for sustainable growth of global nuclear energy in the current century. The GAINS project outlined a framework that provides common platform for methodological and dynamic assessment of global NES covering basic assumptions and boundary conditions. The project also performed sample studies and identified potential areas of GAINS framework application for assessment of key scenarios of transition to sustainable future of nuclear energy systems. Global and regional scenario studies have shown prospects of innovative NES employing closed fuel cycle and containing fast reactors for meeting global and regional nuclear power demands. An INPRO study on thorium fuel cycle has illustrated that thorium can play an important role in supporting U-Pu based NFC for several scenarios considering high demand growth of world nuclear energy. A joint study performed assessment of a national NES based on closed nuclear fuel cycle with fast reactors for determining milestones for deployment and establishing frameworks and areas of collaboration for sustainable nuclear energy development. Although INPRO studies encompass broad areas of technological options for supporting transition to sustainable NES, there are certain possible scenarios that are not considered in these studies. For example, the nuclear reactors used in these studies do not include all reactor design options being developed or considered by Member States, such as high conversion thermal reactors. Similarly, innovative small modular reactor designs with high conversion ratios are at advanced design stages or prototype deployment stage. Such high conversion reactor options can help reduce natural uranium consumption if considered as replacement of conventional LWRs and HWRs, significantly affecting the requirements of FRs or closed nuclear fuel cycle options. Another interesting option is use of FRs with enriched uranium as start-up load followed by utilizing recycled Pu+U+MA from own spent fuel. This option is briefly discussed. The present study provides detailed analysis and investigation of the likely impacts of FR introduction using UOX fuel as startup load on the nuclear energy system.