[en] The Multi-purpose Hybrid Research Reactor for High-tech Applications (MYRRHA) is the flexible experimental accelerator-driven system (ADS) in development at SCK.CEN in replacement of the materials testing reactor BR2 of SCK.CEN in Mol (Belgium). The Belgian federal government recently approved the funding for this international project, which from 2023 onwards will contribute to the development of innovative solutions in the field of nuclear technologies. The coupling between an accelerator, a spallation target and a subcritical core has been studied for the first time at SCK.CEN in collaboration with Ion Beam Applications (IBA, Louvain-la-Neuve) in the framework of the ADONIS project (1995-1997). ADONIS was a small irradiation facility based on the ADS concept, having a dedicated objective to produce radioisotopes for medical purposes and more particularly ⁹⁹Mo as a fission product from highly-enriched ²³⁵U (HEU) fissile targets. The ad hoc scientific advisory committee recommended extending the purpose of the ADONIS machine to become a materials testing reactor (MTR) for materials and fuel research, to study the feasibility of minor actinide transmutation and to demonstrate at a reasonable power scale the principle of the ADS. The project, named MYRRHA as of 1998, then evolved to a larger installation. MYRRHA is now conceived as a flexible irradiation facility, able to work as accelerator-driven (subcritical mode) and in critical mode. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for Generation IV systems, material developments for fusion reactors, radioisotope production for medical and industrial applications and industrial applications, such as Si-doping. MYRRHA will also demonstrate the ADS full concept by coupling the three components (accelerator, spallation target and subcritical reactor) at reasonable power level to allow operation feedback, scalable to an industrial demonstrator and allow the study of efficient transmutation of high-level nuclear waste. Since MYRRHA is based on heavy liquid metal technology (lead-bismuth eutectic), it will be able to significantly contribute to the development of lead fast reactor technology and in critical mode, MYRRHA will play the role of European Technology Pilot Plant in the road-map for LFR. In this paper the historical evolution of MYRRHA and the rationale behind the design choices are presented and the latest configuration of the reactor core and primary system is described. (authors)

[en] Document in abstract form only. Full text of publication follows: Concerns about environmental preservation have increased the demand for more efficient management, as well as sound and sustainable development, of nuclear energy. Appropriate management of radioactive waste arising at the back end of the fuel cycle is considered to be a crucial issue of long-term environmental concern in relation to the nuclear fuel cycle. Of particular concern - due to long-term radiotoxicity and heat load issues, as well as proliferation risk - is the approximately 0.4 wt.% of spent fuel composed of minor actinide (MA) isotopes. As of the year 2006, it is estimated that about 110 tonnes of MA are being contained in spent fuel storage worldwide, and an additional 40 tonnes are contained in high-level waste products from reprocessing for the future. P and T will be necessary to decrease these amounts. Within this framework, this paper presents recent progress obtained at CEA/Marcoule on the development of innovative MA partitioning hydrometallurgical processes in support of their recycling, either in homogeneous mode (MA are recycled at low concentration in all the standard reactor fuel) or in heterogeneous mode (MA are recycled at higher concentration in specific targets, at the periphery of the reactor core). Recovery performances obtained on recent tests under high active conditions of the GANEX process (grouped actinide separation connected to homogeneous recycling) are presented and discussed, as compared to the demands of P and T scenarios. New results also concern major improvements and possible simplifications of the DIAMEX-SANEX process, whose technical feasibility was already demonstrated in 2005 for americium and curium partitioning (heterogeneous mode). The promising initial results of the EXAm process (extraction of only Am, at a front head step of the partitioning process) are also presented. Since recycling of only the americium could be more easily implemented than the recycling of both americium and curium, EXAm is dedicated to Am recovery at an early stage of the partitioning process, using one hydrometallurgical cycle. The principle of the EXAm process is based on the extraction of americium together with some light lanthanides having close values of distribution coefficients in high nitric acidity, and curium and other lanthanides remaining in the aqueous phase. The TEDGA amide molecule is used in order to increase the selectivities Am/Cm and Am/heavy Ln, because of the complexation of curium and heavy lanthanides by this amide; the overall efficiency of the process is largely improved, with a corresponding decrease in the number of necessary separation stages. A test of this innovative process was carried out at the CEA Atalante facility in 2009, with very satisfactory results, and the concentrations of the radioelements measured during the test are in good agreement with the values calculated with the PAREX code. A complete hot test, from a genuine concentrated spent fuel dissolution solution, is planned in 2011-2012, in order to confirm these promising results. In the coming years, subsequent steps will involve both better in-depth understanding of the scientific basis of these actinide recycling processes, and for the new, promising concepts, the studies necessary prior to potential industrial implementation. (authors)

[en] The Czech R and D programme in the field of partitioning and transmutation is grounded on the molten salt reactor system concept with fluoride-salt-based liquid fuel, the fuel cycle of which is based on pyrochemical fluoride partitioning of spent fuel. Two main fluoride partitioning technologies applicable within the MSR fuel cycle are under development at the Nuclear Research Institute Rez plc. The first technology devoted to the reprocessing of LWR or FR spent fuel and to the primary processing of MSR transuranium fuel is the fluoride volatility method. The second technology under development is an electrochemical separation process from fluoride molten salt media. The electrochemical separation should be mainly used for 'on-line' reprocessing of MSR fuel. R and D on the fluoride volatility method is focused on the development and experimental verification of a semi-pilot technology for reprocessing of current and advanced types of oxide spent fuels from LWR or FR. The technology is based on direct fluorination of powdered spent fuel with fluorine gas and on subsequent separation of fluorinated products based on the differences in their volatility. R and D on electro-separation processes from fluoride molten salt media is focused on the development of a suitable electro-separation technique for partitioning of actinides from fission products in the fluoride melt media. The paper summarises the results achieved in the development of pyrochemical partitioning technologies mentioned above and outlines future activities in the Czech P and T programme. (authors)

[en] For more than 20 years, the OECD Nuclear Energy Agency (NEA) has been conducting international studies related to the partitioning and transmutation (P and T) of nuclear waste, including physics, chemistry and material issues, as well as fuel cycle and radioactive waste management strategies. The information exchange meetings on actinide and fission product partitioning and transmutation, organised under the auspices of the Nuclear Science Committee (NSC) and the Nuclear Development Committee (NDC), are part of this programme with the objective of enhancing the value of basic research in the field of P and T by providing experts with a forum to present and discuss current developments in the field and by supporting international collaborations. Recently, the NEA reviewed the potential benefits and impacts of advanced fuel cycles with partitioning and transmutation, as well as national programmes in chemical partitioning. The NEA has also compared criteria for a choice between homogeneous and heterogeneous recycle modes and studied the impact of P and T on potential radiological storage and repositories. This paper provides an overview of results from recent NEA activities in the field of P and T and provides an insight into ongoing projects and planned future activities. (authors)

[en] Document in abstract form only. Full text of publication follows: The YALINA-Booster programme was designed within the framework of the IP-EUROTRANS for validation of the reactivity monitoring techniques on ADS systems. The main objective was the development of a validated methodology to be used in future power ADS. For this purpose, YALINA-Booster is a zero-power subcritical assembly with a fast-thermal neutron spectrum coupled to a D-T neutron generator. The high intensity of the accelerator and the possibility to work in continuous or pulsed mode allowed performing standard pulsed neutron experiments, exploring the current-to-flux relationship and performing beam trip experiments. In addition, it has provided the opportunity to test the electronic chains in current mode, corresponding to the most probable operation mode in a power ADS. Pulsed neutron source (PNS) experiments, previously validated in experiments as MUSE-4, have been carried out to achieve the reference reactivity values for each configuration studied. The techniques applied were the Sjoestrand area method and the prompt neutron decay slope fitting technique. The results obtained from both methods were affected by spatial dependences due to the assembly heterogeneity. These spatial dependences have been corrected using a novel technique based on Monte Carlo simulations. For the first time, the absolute reactivity values of a subcritical system were determined by imposing short millisecond-scale interruptions to the continuous deuterium beam current (beam trips). This technique provided the possibility to monitor the reactivity values on each second using the source jerk methodology. Also for the first time, the fast evolution of the neutron flux intensity within the subcritical assembly was measured using fission chambers operating in current mode. This required the development of specific electronic chains specifically designed for this experimental campaign. In order to test the validity of the experimental results, these reactivity values were demonstrated to be compatible with those obtained by standard PNS techniques. Finally, on-line monitoring of the reactivity has been achieved using the current-to-flux technique. This method relies on the continuous monitoring of the deuteron accelerator current, the neutron source intensity and the neutron flux within the reactor, which is not guaranteed to be proportional to the beam current. With this methodology, relative changes in the subcritical assembly behaviour were possible to be identified in time intervals as short as one millisecond. (authors)

[en] Accuracy of nuclear data, especially for minor actinides (MA), is important in neutronics design of advanced reactors for MA transmutation such as an accelerator-driven system (ADS). A benchmark activity to understand current accuracy of neutronic calculations for the ADS was performed in the Co-ordinated Research Project (CRP) on 'Analytical and Experimental Benchmark Analyses of Accelerator-driven Systems' held by the International Atomic Energy Agency (IAEA). A commercial-grade ADS with a thermal output of 800 MW was employed in the benchmark and depletion analysis was performed by participants using deterministic or Monte Carlo codes. Results revealed that a discrepancy among k-effective by different nuclear data libraries is as large as 2-3 %dk even for an initial criticality before burn-up. Further investigation of the uncertainty was performed using the covariance data in JENDL-3.3 and JENDL-4.0. However, the uncertainty of criticality estimated from JENDL-4.0 was only 1 090 pcm (1.1 %dk), which is much smaller than the previous result of the IAEA-CRP. The comparison of covariance data and cross-section differences among the libraries is necessary. The dominant reactions for uncertainty of criticality, void and Doppler reactivity were identified. The capture cross-sections of ²³⁷Np, ²⁴¹Am and ²⁴³Am were included in the dominant reactions. In order to reduce the uncertainty due to such MA, the Transmutation Physics Experimental Facility (TEF-P) with a significant amount of MA is being proposed in the J-PARC project. An expected reduction of the uncertainty by the TEF-P was numerically evaluated using a cross-section adjustment procedure with covariance data from JENDL-3.3. The original uncertainty for k-effective was about 1.3%. The uncertainty would be reduced from 1.3% to 0.6% if experimental results of the TEF-P and other past critical experiments like the ZPPR were utilised to adjust nuclear data. Therefore, the MA-loaded critical experiment at the TEF-P is important to improve the uncertainty for MA transmutation systems. (authors)

[en] The EURATOM research and training programme for partitioning and transmutation is discussed. The EU has been supporting various P and T related activities since 2007 under the auspices of the 7. Framework Programme (FP7). The European Sustainable Nuclear Energy Technology Platform (SNE-TP) was put in place to help make decisions related to demonstration facilities to be built around a 2015-2020 time horizon. Two collaborative projects, ACSEPT, which studies partitioning technologies and actinide science, and EUROTRANS, which studies transmutation of high-level nuclear waste in ADS, are summarised. (authors)

[en] The Deep Burn (DB) project is a United States Department of Energy (DOE) feasibility study of transuranic management using high burn-up fuels in the high-temperature reactor (HTR). The project started in October 2008 and it is currently in its second year. The project is carried out by a team consisting of the Idaho National Laboratory (INL); Oak Ridge National Laboratory (ORNL); Argonne National Laboratory (ANL); Los Alamos National Laboratory (LANL); several universities, and other private entities. The objective of the project is to establish the technology foundations needed to evaluate the role of the HTR in the nuclear fuel cycle, by providing the information needed to: evaluate cost-effective and economically attractive recycle options for light water reactor (LWR) fuel using the HTR high burn-up capabilities that will reduce transuranic (TRU) stockpiles; evaluate the capacity of geologic facilities to store HTR spent fuel, both low-enriched uranium (LEU) and DB fuel; consider the HTR as an important part of nuclear growth scenarios that include closing the fuel cycle; incorporate HTR fuel cycle issues into the mission of the next generation nuclear plant (NGNP). Highlights of activities and achievements of the project will be presented, including: 1. Core and fuel analysis. The neutron energy spectrum in a DB-HTR differs from that in the ordinary HTR, which is based on the use of enriched uranium as the fissile material in the fuel. Therefore, modifications to the existing HTR core physics modelling tools had to be implemented. Analyses of both the prismatic and pebble-bed core types have been carried out that quantify the TRU destruction capability, the power and fuel temperature peaking in the core, and the temperature reactivity coefficients. The performance of the coated particle fuel is quantified in terms of the maximum stress in the silicon carbide coating layer and in terms of its potential consequent failure probability during irradiation in the core. The initial conceptual core designs have been optimised to retain acceptable predefined safety limits together with a high TRU destruction performance in excess of 60%. 2. Spent fuel management. The basic models required to determine radionuclide release from spent TRISO coated fuel were defined. 3. Fuel cycle integration. The DANESS code has been used to do preliminary modelling of the various ways that LWR, HTR and sodium fast reactors (SFR) can be combined to reduce TRU waste (synergy scenarios). 4. Fuel modelling. Recent advances in the understanding of materials behaviour and computing capability are used to enhance analytical modelling for high burn-up coated particle fuel. The task is divided into three sub-tasks: thermochemical modelling, transport phenomena and radiation damage. 5. Fuel qualification. Progress has been made in laboratory-scale development of TRU TRISO-coated particle fuel, including kernel fabrication, fluidized bed coating, kernel and coated particle characterisation and analysis of ZrC material properties. 6. Spent fuel recycle. The recycle of HTR spent fuel is being investigated in the project to address the needs of LWR TRU recycle options under consideration (specifically the synergy scenarios), and to provide integration of the HTR into the fuel cycles investigated by the Fuel Cycle Research and Development (FCRD) programme. (authors)

[en] Document in abstract form only. Full text of publication follows: Innovative nuclear systems cooled with liquid Pb and Pb-Bi eutectic (LBE) such as e.g. the accelerator-driven transmutation system or the lead fast reactor can be potentially used for the transmutation of high-level nuclear wastes. The safe operation of these systems implies, among other items, a detailed knowledge of the materials' performance in the liquid metal environment and the handling of the liquid metal in terms of soluble and insoluble impurities. The objective of this contribution is to summarise relevant studies conducted recently in Europe on materials selection and characterisation in terms of corrosion and mechanical resistance in the liquid metal. The analysis will be done by taking into account the importance of the HLM chemistry. In particular the oxygen effect will be illustrated and techniques able to adjust and measure the oxygen content in the liquid metal will be described. As far as the selection of the structural materials for the core components, vessel and in-vessel components of liquid-metal-cooled systems, this selection is dictated by several factors such as temperature, mechanical stresses, irradiation damage and corrosion resistance. The corrosiveness of the liquid Pb and LBE has been thoroughly investigated and it has been found that liquid metal chemistry, temperature and flow rate have an important role in the appearance of the corrosion mechanism and in the estimation of the corrosion rate. In particular the oxygen dissolved in the liquid metal, in combination with the different temperature ranges, has the most important effect on the steel corrosion mechanism. Indeed, steels containing elements highly soluble in the liquid metals as, e.g. the class of austenitic steels, are oxidised when the level of oxygen in the liquid reaches adequate potentials and the oxidation rate does not exceed tenths of micrometers per year. However, even under oxidising conditions the same steel suffers severe dissolution corrosion for temperatures above ∼450 deg. C. Dissolution attack depends on the temperature and it can reach even hundreds of micrometers of penetration from the surface to the bulk. The consequences of severe dissolution of austenitic steels is a change in the microstructure of this material, since Ni is leached out (due to the highest solubility in the Pb and LBE) and the austenitic structure changes to ferritic until other elements such as Cr and Fe are also dissolved. This mechanism can induce a reduction of the strength, i.e. the load-bearing capability of the steel. On the other hand the class of ferritic/martensitic steels has a different behaviour. Indeed, these steels are oxidised with a considerable rate in the high-temperature range (∼500-550 deg. C). A high oxidation rate implies a thick oxide layer which can spall off and allow liquid metal penetration with consequent change of corrosion mechanism from oxidation to dissolution. These events must be avoided, since ferritic/martensitic steels suffer mechanical properties degradation due to the liquid metal. However, a proper oxygen potential control in the liquid metal and for lower temperatures (< 500 deg. C) the oxide layer may act as corrosion protection barrier and the thickness of the scale may stay within acceptable levels. The control, in terms of measurement and setting, of the oxygen potential in the liquid metal can occur with dedicated techniques. Indeed, oxygen potential measurement can be done with electrochemical probes and the adjustment of the oxygen level can be performed through the gas phase or mass exchange devices. Electrochemical probes with different reference electrodes have been developed and it turned out that the Pt/air reference electrode is very reliable for oxygen potential measurement in the high-temperature range, while Me/MeO (Me: Bi, In, etc.) are reference electrodes that can be used for measurement in lower temperature ranges. As stated above, the objective of this contribution is to illustrate the relation be tween structural materials' performance and liquid metal chemistry and the impact of these two items on design and safety aspects of nuclear transmutation reactors cooled with HLM. (authors)

[en] Partitioning and transmutation (P and T) is one of the key technologies for reducing the radiotoxicity and volume of radioactive waste arisings. Recent developments indicate the need for embedding P and T strategies in advanced fuel cycles considering both waste management and economic issues. In order to provide experts a forum to present and discuss state-of-the-art developments in the P and T field, the OECD/NEA has been organising biennial information exchange meetings on actinide and fission product partitioning and transmutation since 1990. The previous meetings were held in Mito (Japan) in 1990, at Argonne (United States) in 1992, in Cadarache (France) in 1994, in Mito (Japan) in 1996, in Mol (Belgium) in 1998, in Madrid (Spain) in 2000, in Jeju (Korea) in 2002, in Las Vegas (United States) in 2004, in Nimes (France) in 2006 and in Mito (Japan) in 2008. They have often been co-sponsored by the European Commission (EC) and the International Atomic Energy Agency (IAEA). The 11. Information Exchange Meeting was held in San Francisco, California, United States on 1-4 November 2010, comprising a plenary session on national P and T programmes and six technical sessions covering various fields of P and T. The meeting was hosted by the Idaho National Laboratory (INL), United States. The information exchange meetings on P and T form an integral part of NEA activities on advanced nuclear fuel cycles. The meeting covered scientific as well as strategic/policy developments in the field of P and T, such as: fuel cycle strategies and transition scenarios; radioactive waste forms; the impact of P and T on geological disposal; radioactive waste management strategies (including secondary wastes); transmutation fuels and targets; pyro and aqueous separation processes; materials, spallation targets and coolants; transmutation physics, experiments and nuclear data; transmutation systems (design, performance and safety); handling and transportation of transmutation fuels; and economics of P and T. These topics encompass subjects such as multi-scale modelling, experiments and instrumentation, novel theories and benchmarks. A plenary session was organised to present national programmes related to P and T. A total of 102 presentations (42 oral and 60 posters) were discussed by 131 participants from 17 countries and 3 international organisations. A closing session with a panel discussion was also organised at the end of the meeting. These proceedings include all the papers presented at the 11. Information Exchange Meeting