1,721,080 research outputs found
Design dell’esperimento e specifica tecnica di fornitura relativa all’up-grade dell’impianto LIFUS5/Mod2
No full textIl documento riguarda la specifica di fornitura dell’impianto LIFUS5 per la campagna sperimentale da effettuare nell’ambito del progetto FP7 EC MAXSIMA. In tale ambito, l’impianto sperimentale sarà modificato e aggiornato al fine 1) di ripristinare la funzionalità dei componenti soggetti a usura durante le campagne EC THINS e EC LEADER, 2) di modificare il layout del sistema di iniezione al fine di consentire l’utilizzazione dell’impianto anche per lo studio delle piccole perdite dai tubi del generatore di vapore, 3) di consentire l’utilizzazione di più fluidi di prova. Inoltre, il sistema di acquisizione e strumentazione sarà aggiornato con l’istallazione di nuova strumentazione. Il risultato sarà la realizzazione di un’apparecchiatura sperimentale “multi-purpose” che sarà denominata LIFUS5/Mod3
Assessment of SIMMER-III code in predicting water cooled lithium lead breeding blanket “in-box-Loss of Coolant Accident”
Full text linkThe in-box Loss of Coolant Accident is a major safety concern for the Water Cooled Lithium Lead Breeding Blanket design. SIMMER-III code has been modified by University of Pisa and ENEA C.R. Brasimone to perform deterministic safety investigation of such accidental scenario. Thanks these modifications, this version of the code has unique features for dealing with the PbLi water reaction related phenomena. The reliability of the code requires a systematic validation activity, carried out applying a standard methodology based on a three-steps procedure and through qualitative and quantitative evaluations. The methodology was applied to the LIFUS5 campaign. Post-test analyses highlighted open issues of test execution and of experimental data, as well as code limitations and capabilities. Nevertheless, the modified version of SIMMER-III code for fusion application is able to predict the relevant thermal-hydraulic phenomena during PbLi/water interaction
La collezione erpetologica del Museo Zoologico dell???Università di Napoli Federico II
No full textPubblicato su: PIANURA: SCIENZE E STORIA DELL'AMBIENTE PADANO vol. 1
Connectivity, centralisation and ‘robustness-yet-fragility’ of interbank networks
Full text linkThis paper studies the effects that connectivity and centralisation have on the response of interbank networks to external shocks that generate phenomena of default contagion. We run numerical simulations of contagion processes on randomly generated networks, characterised by different degrees of density and centralisation. Our main findings show that the degree of robustness-yet-fragility of a network grows progressively with both its degree of density or centralisation, although at different paces. We also find that sparse and decentralised interbank networks are generally resilient to small shocks, contrary to what so far believed. The degree of robustness-yet-fragility of an interbank network determines its propensity to generate a too-many-to-fail problem. We argue that medium levels of density and high levels of centralisation prevent the emergence of a too-many-to-fail issue for small and medium shocks whilst drastically creating the problem in the case of large shocks. Finally, our results shed some light on the actual robustness-yet-fragility of the observed core-periphery national interbank networks, highlighting the existing risk of systemic crises
DEVELOPMENT AND VALIDATION OF A SIMMER-III/ANSYS CODE CHAIN METHODOLOGY FOR THE INTEGRAL SAFETY ANALYSIS OF WCLL FUSION REACTOR COMPONENTS
No full textIn the framework of the development of fusion energy, one of the most prominent technologies arising to address the issues of tritium breeding and power conversion is the Water-Cooled Lithium-Lead (WCLL). This technology utilizes a molten eutectic alloy of Lithium and Lead which circulates inside the Breeding Blankets (BB) and is irradiated with neutrons to produce tritium. Water is then circulated inside the system to cool the components. The simultaneous presence inside critical areas of the reactor of molten metal alloys and water, at high temperature and pressure, poses significant safety concerns. For this reason, adequate design and analysis techniques are required to ensure the ability of the system to survive and mitigate any possible damage in case of the in-box Loss Of Coolant Accident (LOCA), the most critical postulated accidental scenario. This work introduces a new methodology for the integral safety analysis of WCLL components, with a particular focus on the WCLL Breeding Blankets, which is based on a fully automated code-chain technique. Its goal is to couple the calculations performed in the fluid domain by the SIMMER-III code, which models the chemical and thermodynamical interactions between the water and the alloy, and the structural simulations performed by the ANSYS code on the mechanical components. The entire process is validated against experimental data provided by the LIFUS5 facility operating at ENEA Brasimone Research Centre. The resulting comparison between these data and codes' predictions allows a careful evaluation of the errors introduced in each step of the chain. Moreover, it provides confidence in the capacity of the methodology to correctly predict the ability of the structures to withstand incidental loads without suffering extensive damage. This work aims at providing engineers with a usable and powerful tool that allows for the safety analysis of WCLL-based components during the early stages of the design phase. This would help save time, and effort and reduce the economic cost that might arise from any undetected issue propagating downstream the design process
La collezione erpetologica del Museo Zoologico dell???Università di Napoli Federico II
No full textPubblicato su: PIANURA: SCIENZE E STORIA DELL'AMBIENTE PADANO vol. 1
Experimental campaign in support of the safety studies of the STGR in LFR
No full textThe Steam Generator Tube Rupture postulated event in a pool type Gen IV Heavy Liquid Metal cooled Fast Reactor system needs to by deeply investigated due to the possible hazardous consequences affecting the structural integrity of internals. These reactor designs have steam generators inside the reactor vessel, thus, the interaction between the secondary side coolant (water) and the HLM (e.g. steam generator tube rupture) has to be considered as challenging safety issue in the design and also in the preliminary safety analysis of these reactor types. The LIFUS5/Mod2 facility experiments series B were executed in the framework of the FP7 EC Lead-cooled European Advanced Demonstration Reactor – LEADER – project. The test section was representative of the Spiral Tube Steam Generator of European Lead Fast Reactor. Water at about 180 bar and 270°C was injected into Lead Bismuth Eutectic alloy at about 2 bar and 400°C. The injection was performed in the center of a tube bundle composed of 188 tubes (a representative portion of SG bundle). This analysis is aimed at providing engineering feedbacks to SG designers (e.g. domino effect and mechanical loadings on components). Moreover, the acquired experimental data constituted a wide database for development and validation of numerical models and calculation codes
Gli Anfibi del Museo Zoologico dell'Università di Napoli Federico II: catalogo della collezione con note storiche. Atti Mus. civ. Stor. Nat., Trieste
No full text
Experimental campaign in support of the safety studies of the STGR in LFR
No full textThe Steam Generator Tube Rupture postulated event in a pool type Gen IV Heavy Liquid Metal cooled Fast Reactor system needs to by deeply investigated due to the possible hazardous consequences affecting the structural integrity of internals. These reactor designs have steam generators inside the reactor vessel, thus, the interaction between the secondary side coolant (water) and the HLM (e.g. steam generator tube rupture) has to be considered as challenging safety issue in the design and also in the preliminary safety analysis of these reactor types. The LIFUS5/Mod2 facility experiments series B were executed in the framework of the FP7 EC Lead-cooled European Advanced Demonstration Reactor – LEADER – project. The test section was representative of the Spiral Tube Steam Generator of European Lead Fast Reactor. Water at about 180 bar and 270°C was injected into Lead Bismuth Eutectic alloy at about 2 bar and 400°C. The injection was performed in the center of a tube bundle composed of 188 tubes (a representative portion of SG bundle). This analysis is aimed at providing engineering feedbacks to SG designers (e.g. domino effect and mechanical loadings on components). Moreover, the acquired experimental data constituted a wide database for development and validation of numerical models and calculation codes. © 2019 American Nuclear Society
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