1,721,053 research outputs found
Validazione su base sperimentale del codice RELAP5 mediante campagna di prove sperimentali su impianto HEFUS-3
Full text linkPresso il centro di ricerca ENEA Brasimone, nell'ambito dell' Accordo di Programma tra Ministero dello Sviluppo Economico ed ENEA, è stata svolta un'attività sperimentale sull'impianto ad elio Hefus3 ai fini di validare il codice di calcolo RELAP5. Le attività sperimentali sono state condotte con portate fino a 900 g/s a 70 bar, in regime di refrigerazione isoterma e a lungo termine, incidente di LOFA (Loss of Flow Accidents) ed incidente di LOCA (Loss of Coolant Accidents). I set di dati generati sono stati impiegati per integrare i dati precedentemente acquisiti con range di portate fino a 0.35 kg/s a 4 MPA per la validazione dei codici T/H per l'analisi di reattori a gas ad alta temperatura mediante codice RELAP5
Facilities, testing program and modeling needs for studying liquid metal magnetohydrodynamic flows in fusion blankets
No full textSince many years, liquid metal flows for applications in fusion blankets have been investigated worldwide. A review is given about modeling requirements and existing experimental facilities for investigations of liquid metal related issues in blankets with the focus on magnetohydrodynamics (MHD). Most of the performed theoretical and experimental works were dedicated to fundamental aspects of MHD flows under very strong magnetic fields as they may occur in generic elements of fusion blankets like pipes, ducts, bends, expansions and contractions. Those experiments are required to progressively validate numerical tools with the purpose of obtaining codes capable to predict MHD flows at fusion relevant parameters in complex blanket geometries, taking into account electrical and thermal coupling between fluid and structural materials. Scaled mock-up experiments support the theoretical activities and help deriving engineering correlations for cases which cannot be calculated with required accuracy up to now. © 2014 Karlsruhe Institute of Technology All rights reserved
The tritium extraction and removal system for the DCLL-DEMO fusion reactor
No full textDuring the pre-conceptual design phase of DEMO, different alternatives have been explored to be implemented as tritium extraction and removal system (TERS) for the blanket concepts considered in EUROfusion. The TERS is conceived to extract tritium from the breeder and to route it to the Tritium Plant for final processing. A careful review showed that those blankets operated with PbLi should use the permeation against vacuum (PAV) technique as primary option which is based on a one-step, fully continuous procedure. In this paper, a conceptual design of the TERS for the dual coolant lithium lead (DCLL) breeding blanket is presented, based on the European DEMO2015 layout (18 sectors, 2037 MW fusion power). The P&ID of the proposed TERS, integrated in the DCLL-PbLi loop, includes valves and instrumentation, as well as a revised design of the DCLL-PAV. The dimensioning of the permeator considered a tritium extraction efficiency of 80%. An exhaustive investigation on the vacuum system needed for the PAV is also presented. The choice of the most promising vacuum systems took into account the reliability and tritium compatibility of both high and rough pumps. Their pumping requirements, which are dependent on the PAV efficiency, tritium solubility and tritium partial pressure in the loop, are also discussed in this work. © EURATOM 2018
Design of a Permeator-Against-Vacuum mock-Up for the tritium extraction from PbLi at low speed
No full textThe Permeator Against Vacuum (PAV) is one of the candidate Tritium extraction and removal technology from the Lithium-Lead circulating in the breeding blanket currently under consideration for the EU DEMO. In this work the design of a PAV mock-up, dimensioned for the test in the TRIEX facility at ENEA Brasimone, is presented, together with a structural and fluid-dynamic analysis. The mock-up consists of parallel helical vanadium pipes inserted in a vacuum vessel, where the Tritium permeates from the PbLi. The extraction efficiency of the mock-up is computed using an established Tritium permeation model, demonstrating that the mock-up reaches the target 80% efficiency in nominal conditions. © 201
Tritium control in fusion reactor materials: A model for Tritium Extracting System
No full textIn fusion reactors, tritium is bred by lithium isotopes inside the blanket and then extracted. However, tritium can contaminate the reactor structures, and can be eventually released into the environment. Tritium in reactor components should therefore be kept under close control throughout the fusion reactor lifetime, bearing in mind the risk of accidents, the need for maintenance and the detritiation of dismantled reactor components before their re-use or disposal. A modeling work has been performed to address these issues in view of its utilization for the TES (Tritium Extraction System), in the case of the HCPB TBM and for a molecular sieve as adsorbent material. A computational model has been setup and tested. The results of experimental measurement of fundamental parameters such as mass transfer coefficients have been implemented in the model. It turns out the capability of the model to describe the extraction process of gaseous tritium compounds and to estimate the breakthrough curves of the two main tritium gaseous species (H2 and HT). © 2015 Elsevier B.V. All rights reserved
Literature review of lead-lithium thermophysical properties
No full textIn recent years, the scientific community of nuclear fusion raised the issue of thermophysical properties of lead lithium alloys. These alloys are foreseen to be used in several Breeding Blanket concepts in an almost eutectic composition, but only few data on the properties are available in literature and large differences on the same property exist between different authors. Moreover, apparently each organization used different available properties correlations, making practically pointless every comparison of results with the other organizations involved in the design of Breeding Blankets. The aim of this paper is to identify the properties to be used in the design of the Breeding Blankets, performing a literature review of the available data and suggesting a correlation for each of the main properties. These correlations were chosen based on the accurateness of the paper and on the similarities between different authors, where it was possible (e.g., density). The table with the correlations should represent a starting point for a discussion to reach a general consensus on the property database, which should be mandatory in order to allow a comparison of the results from different organizations. Very likely new experiments will be necessary to definitely measure at least the properties with the biggest scattering of the data (e.g., specific heat), encouraging a consensus and reducing the errors in the design activities. © 201
Radiation effects on deuterium permeation for PLD alumina coated Eurofer steel measured during 1.8 MeV electron irradiation
No full textTritium permeation through structural materials in future fusion reactors is an extremely important issue due to both radiological concerns and tritium self-sufficiency. In the past many efforts to minimize tritium losses have been made and permeation barriers are considered the best method to reduce them. Alumina is one of the main candidates for such barriers due to its chemical stability and hydrogen isotope permeation reduction factor. In this work pulsed laser deposited (PLD) alumina on Eurofer steel has been studied in terms of deuterium permeation. A reduction factor of about 1000 compared to bare material in the temperature range RT-450 °C was measured. In order to study deuterium permeation during irradiation the Radiation Induced Permeation and Release (RIPER) facility at Ciemat was used. Contrary to expectations the permeation through PLD alumina coated Eurofer was observed to decrease during irradiation. This radiation effect was explained making use of thermostimulated desorption (TSD) experiments. TSD measurements were carried out for both electron irradiated and unirradiated alumina samples and the results showed that radiation induced deuterium back desorption from alumina occurred in this way reducing the overall permeation value during irradiation. Finally, the microstructural study of the irradiated coatings showed that some surface damage appears but not cracking or coating debonding was observed in agreement with the good behaviour observed in terms of deuterium permeation for the irradiated coatings. © 2018 Elsevier B.V
Efficient hydrogen and deuterium permeation reduction in Al2O3 coatings with enhanced radiation tolerance and corrosion resistance
No full textOne of the major bottlenecks in the development of the breeding blanket of the DEMO fusion reactor is the suppression of tritium permeation through structural steels as well as their protection against dissolution-corrosion by interaction with high temperature heavy liquid metals. Reduction of tritium permeation and corrosion of structural steels are crucial issues in order to enhance reactor safety and avoid operational implications. As a solution to these two daunting challenges, we developed multifunctional alumina coatings capable to tackle, at the same time, tritium permeation and Pb-Li corrosion. The coatings are deposited by pulsed laser deposition and are essentially amorphous, with nanocrystalline inclusions. By optimizing the deposition process, we provide experimental evidence that PLD-grown alumina yields a permeation reduction factor for hydrogen and deuterium well above 104, even after electron irradiation, and a suitable protection against corrosion by the Pb-16Li eutectic. Given these results, the multifunctional, PLD-grown alumina coatings stand out as a viable solution to some of the long-lasting issues related to fusion technologies. © 2018 IAEA, Vienna
Finalization of the conceptual design of the auxiliary circuits for the European test blanket systems
No full textIn view of the ITER conceptual design review, the design of the ancillary systems of the European test blanket systems presented in [1] has been updated and made consistent with the ITER requirements for the present design phase. Europe is developing two concepts of TBM, the helium cooled lithium lead (HCLL) and the helium cooled pebble bed (HCPB) one, having in common the cooling media, pressurized helium at 8 MPa [2]. TBS, namely helium cooling system (HCS), coolant purification system (CPS), lead lithium loop and tritium extraction/removal system (TES-TRS) have the purpose to cool down the TBM and to remove tritium to be driven to TEP from breeder and coolant. These systems are placed in port cell 16 (PC#16), chemical and volume control system (CVCS) area and tritium building. Starting from the pre-conceptual design developed in the past, more mature technical interfaces with the ITER facility have been consolidated and iterative design activities were performed to comply with design requirements/specifications requested by IO to conclude the conceptual design phase. In this paper the present status of design of the TBS is presented together with the preliminary integration in ITER areas. © 2015 Elsevier B.V
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