[en] The reduction of tritium permeation through blanket structural materials and cooling tubes has to be carefully evaluated to minimise radiological hazards. A strong effort has been made in the past to select the best technological solution for the realisation of tritium permeation barriers (TPB) on complex structures not directly accessible after the completion of the manufacturing process. The best solution was identified in aluminium rich coatings, which form Al₂O₃ at their surface. Two technologies were selected as reference for the realisation of coating in the WCLL blanket concept: the chemical vapour deposition (CVD) process developed on laboratory scale by CEA, and the hot dipping (HD) process developed by FZK. The results obtained during three years of tests on CVD and HD coated specimens in gas and liquid metal phase are summarised and discussed

[en] The interaction between pressurised water and liquid Pb-17Li is a topical issue for the WCLL blanket concept and needs to be studied in detail because of the potentially significant effect on the reliability and safety of the blanket. Particularly, it seems important from the design point of view to deeply investigate the consequences of coolant micro-leaks in the blanket module generated by micro-cracks with a size in the order of 10^-3 mm². This kind of interaction has been studied during an extensive experimental campaign on the RELA loops in the ENEA Research Centre at Brasimone. The results showed a reduction in the liquid metal flow-rate in the circuit and a deterioration of the heat exchange properties between breeder and coolant due to the formation and growth of solid reaction products generated by the chemical reaction between lithium and water. The molar ratio between the recovered hydrogen recovered and water injected was in the range 0.3-0.5, confirming that, in the test conditions, the main solid reaction product is lithium hydroxide (LiOH). The possible impact of these results on the TBM-ITER design is also presented and discussed

[en] In the helium cooled lithium lead (HCLL) and water cooled lithium lead (WCLL) blanket concepts for DEMO correct and reliable management of tritium is of basic importance, both for safety and fuel cycle reasons. To develop a sensor for measurements of hydrogen (and its isotopes) concentration in liquid Pb-17Li, a permeable capsule of niobium was chosen. Different simulations with a mathematical model have been performed, and then the sensor was designed, constructed and tested. The first experimental results in gas phase showed a permeating flux much lower than the predicted one, probably due to the formation of an oxide layer on the capsule surface or to the formation of niobium hydrides. To solve this problem different solutions are presented

[en] The general methodology of advanced thermo-hydraulic calculations defined in this article has been developed, based on preliminary design of a water cooling circuit in a fusion reactor, to continue with the design process to ensure the justification and optimization of the design of the different subsystems with the ultimate purpose of achieving the final design of the system.

[en] Since several years ENEA is involved in experimental activities concerning the interaction between molten lithium lead alloy, in eutectic composition, and pressurised water in conditions relevant for DEMO fusion reactor, in order to predict the behaviour of a WCLL blanket module in case of an in blanket LOCA. In this ambit LIFUS 5 apparatus was designed and constructed to carry out the experimental campaign on water large leaks into liquid Pb-17Li, with the final aim to provide data for the validation of the mathematical modelling of the related phenomena. After two tests performed to qualify all mechanical components and the data acquisition system, LIFUS 5 was operated at the ENEA C.R. Brasimone for tests nos. 3 and 4. Water was injected into the reaction tank at a pressure of 155 bar with two different values of sub-cooling. The initial liquid metal temperature was fixed to 330 deg. C. The first pressure peak due to the water vaporization and jet expansion was clearly recognized together with the subsequent pressure increase due to further water injection and hydrogen generation. In the performed experiments the maximum pressure peak, as detected in both reaction and expansion vessels, never overcame the value of the injected water. Moreover, in these two tests a significant temperature increase in the reaction vessel occurred, strictly connected to the amount of injected water

[en] The reduction of tritium permeation from the Pb-17Li, or plasma, into the coolant is of crucial importance in order to reduce the radiological hazard in the steam generator vault as well as in the turbine/condenser area and to optimise the tritium balance in the reactor. The use of aluminium rich coatings has been selected as reference solution for the water cooled lithium lead (WCLL) blanket in order to produce reliable tritium permeation barriers (TPB). TPB qualification activities performed in the past allowed the selection of two reference deposition techniques, the chemical vapour deposition (CVD) process developed on laboratory scale by CEA, and the hot dipping (HD) process developed by FZK. On the basis of the results obtained in the past with the Corelli I-II devices, a new apparatus named Vivaldi was designed to perform comparative tests on two hollow cylindrical specimens in the same operating conditions. The performance of alumina coating on EUROFER 97 steel has been tested in gas and liquid metal phase. The obtained results in terms of permeated fluxes and permeation reduction factors (PRF) are herein presented and discussed. A post experiment examination of coatings was performed by use of optical and SEM microscopy

[en] Ignitor is a nuclear fusion reactor aimed at studying Deuterium-Tritium plasmas. We have applied the Italian waste management regulations to the IGNITOR experiment radioactive materials: none of them should be classified in the high level waste category and most materials are classified as low level waste. If European proposed waste management strategies were applied, all Ignitor radioactive materials could be recycled

[en] Chronical releases of tritium from the helium primary coolant into the water secondary coolant is a fundamental safety issue in the design of a fusion reactor steam generator. It is well known that the steam/water circuit of a fusion reactor would be considered not relevant from a radiological point of view, while if a strong permeation of tritium will be present it will be released together with incondensable gases in the condenser. The permeation of hydrogen isotopes through candidate steam generator materials in different conditions was studied in the past. Further experiments demonstrated that nickel alloys of nuclear interest are always covered by a thin and adherent oxide layer able to reduce permeation of orders of magnitude. The major objective of this work is the evaluation of the permeated flux through nickel alloys, when exposed to pure hydrogen and to an oxidant gas stream, to verify the real permeability of these materials in conditions close to those foreseen in the helium side of the steam generator.

[en] Conclusions: • The ENEA objective is to support the Italian industrial system in the design, development and construction of the LFR demonstrator ALFRED and, in a longer time, of the ELFR First Of A Kind. • The HLM technology is under development in several European countries. • The viable route to ALFRED seems to be the construction in a European country able to access the EC structural funds for cohesion: Romania has already expressed this purpose. • The conceptual design is still ongoing. The numerical tools for heavy liquid metals have been implemented and tested since years. • The R&D associated with main components is ongoing. Most of the necessary facilities are operational. • The design trade-off between a “adiabatic core” and a “safe core” is achieved. • The issue of fuel cladding material remains the toughest issue to deal with. This is mainly due to the qualification effort which must be envisaged even in irradiation facilities. • Italy is looking for a foreign partner for irradiations aimed to LFR fuel development testing

[en] Since the end of the 1990s, ENEA Brasimone Research Centre operates several experimental facilities aimed at supporting the research on LBE and lead systems, such as ADS and LFR. The experimental activities are related to thermal- hydraulics, heat exchange, science of materials, and qualification of components. Experimental loops operated at ENEA and an experiment addressing the compatibility of structural materials to withstand with aggressive coolants (i.e. Lead and LBE), at high temperature and high dose rate are described in this paper. These loops have been also operated for supporting the research related to the selection of the cladding materials for ADS and LRF technologies. The paper provides an overview of the recent and future activities devoted to material development. (author)