0000000000108512
AUTHOR
Pietro Alessandro Di Maio
showing 108 related works from this author
Safety analysis of the DONES primary heat removal system
2020
Abstract The development of a neutron source able to reproduce the irradiation conditions typical of a nuclear fusion reactor, in order to test candidate structural materials, is the main goal of the Work Package Early Neutron Source (WPENS) of the EUROfusion action. This source, named Demo Oriented NEutron Source (DONES), is a facility where neutrons are produced by means of D-Li interactions. More in detail, a beam of 125 mA deuterium ions at the energy of 40 MeV strikes a lithium jet flowing in a purposely shaped channel in order to obtain an intense and stable neutron flux for the irradiation of material samples. In the framework of these activities, safety analyses are a key aspect in …
Conceptual design of the main Ancillary Systems of the ITER Water Cooled Lithium Lead Test Blanket System
2021
Abstract The Water Cooled Lithium Lead Test Blanket System (WCLL TBS) is one of the EU Test Blanket Systems candidate for being installed and operated in ITER. In view of its Conceptual Design Review by F4E and ITER Organization (IO), planned for mid-September 2020, several technical activities have been performed in the areas of WCLL TBS Ancillary Systems design. In this article the outcomes of the conceptual design phase of the four main Ancillary Systems of WCLL TBS, namely the Water Cooling System (WCS), the Coolant Purification System (CPS), the PbLi loop and the Tritium Extraction System (TES), are reported and critically discussed. In particular, for each Ancillary System hereafter a…
On the effects of the Double-Walled Tubes lay-out on the DEMO WCLL breeding blanket module thermal behavior
2019
Abstract The EU-DEMO Water-Cooled Lithium Lead Breeding Blanket (WCLL BB) concept foresees liquid Pb-15.7Li eutectic alloy as breeder and neutron multiplier, whereas pressurized subcooled water as coolant, with operative conditions typical of the PWR fission reactors (temperature in the range of 295–328 °C and pressure of 15.5 MPa). The cooling down of the BB is guaranteed by means of two separated cooling circuits: the one consisted in square channels housed within the complex of Side Walls and First Wall, and the one composed of a set of Double-Walled Tubes (DWTs) submerged in the Breeding Zone (BZ) and deputed to remove heat power therein generated. A parametric thermal study has been ca…
Parametric thermal analysis for the optimization of Double Walled Tubes layout in the Water Cooled Lithium Lead inboard blanket of DEMO fusion reactor
2019
Abstract Within the roadmap that will lead to the nuclear fusion exploitation for electric energy generation, the construction of a DEMOnstration (DEMO) reactor is, probably, the most important milestone to be reached since it will demonstrate the technological feasibility and economic competitiveness of an industrial-scale nuclear fusion reactor. In order to reach this goal, several European universities and research centres have joined their efforts in the EUROfusion action, funded by HORIZON 2020 UE programme. Within the framework of EUROfusion research activities, ENEA and University of Palermo are involved in the design of the Water-Cooled Lithium Lead Breeding Blanket (WCLL BB), that …
Thermo-Mechanical Analysis and Design Update of the Top Cap Region of the DEMO Water-Cooled Lithium Lead Central Outboard Blanket Segment
2022
Within the framework of the EUROfusion research and development activities, the Water-Cooled Lithium Lead (WCLL) Breeding Blanket (BB) is one of the two candidates to be chosen as the driver blanket for the European DEMO nuclear fusion reactor. Hence, an intense research work is currently ongoing throughout the EU to develop a robust conceptual design able to fulfil the design requirements selected at the end of the DEMO pre-conceptual design phase. In this work, the thermo-mechanical analysis and the design update of the top cap (TC) region of the DEMO WCLL Central Out-board Blanket (COB) segment is presented. The scope of the work is to find a design solution of the WCLL COB TC region abl…
Hydraulic Characterization of the Full Scale Mock-Up of the DEMO Divertor Outer Vertical Target
2021
In the frame of the pre-conceptual design activities of the DEMO work package DIV-1 “Divertor Cassette Design and Integration” of the EUROfusion program, a mock-up of the divertor outer vertical target (OVT) was built, mainly in order to: (i) demonstrate the technical feasibility of manufacturing procedures; (ii) verify the hydraulic design and its capability to ensure a uniform and proper cooling for the plasma facing units (PFUs) with an acceptable pressure drop; and (iii) experimentally validate the computational fluid-dynamic (CFD) model developed by the University of Palermo. In this context, a research campaign was jointly carried out by the University of Palermo and ENEA to experimen…
Structural assessment of the EU-DEMO WCLL Central Outboard Blanket segment under normal and off-normal operating conditions
2021
Abstract Within the framework of the EUROfusion design activities concerning the EU-DEMO Breeding Blanket (BB) system, a research campaign has been carried out at the University of Palermo with the aim of investigating the structural behaviour of the DEMO Water-Cooled Lithium Lead (WCLL) Central Outboard Blanket (COB) segment. The assessment has been performed considering three different loading scenarios: the Normal Operation (NO), the Over-Pressurization (OP) and the Upward Vertical Displacement Event (VDE-up). In particular, NO scenario represents the loading case referring to the nominal operating conditions, whereas the OP scenario refers to the loading conditions due to an in-box LOCA…
Further improvements in the structural analysis of DEMO Divertor Cassette body and design assessment according to RCC-MRx
2019
Abstract This paper presents the enhancements related to the structural analyses of DEMO Divertor in the framework of the EUROfusion Power Plant Physics & Technology (PPPT) program. This activity started two years ago and its preliminary results were published in previous papers. It has been divided in some areas defined by the similarity of the matters they contain: the structural analysis, of utmost importance, has been preceded by a preliminary phase, like the geometry definition or the thermal and the electric-magnetic analysis for loads evaluation; then the structural analysis has been finally confirmed with further evaluations related to excessive deformation or plastic instability. T…
Parametric study of the influence of First Wall cooling water on the Water Cooled Lithium Lead Breeding Blanket nuclear response
2019
Abstract In the framework of EUROfusion Work Package International Cooperation R&D activities, a close collaboration has been started among University of Palermo, ENEA Brasimone and ENEA Frascati for the development of the Water Cooled Lithium Lead (WCLL) Breeding Blanket (BB) concept. In this context, a research campaign has been carried out at the University of Palermo in order to investigate the influence of First Wall (FW) cooling water configuration on the nuclear response of the WCLL BB under irradiation in EU-DEMO, in order to gain useful indications for the WCLL BB pre-conceptual designs. To this end, three-dimensional nuclear analyses have been performed by MCNP5 v. 1.6 Monte Carlo…
Preliminary thermal optimization and investigation of the overall structural behaviour of the EU-DEMO water-cooled lead lithium left outboard blanket…
2022
The conceptual design phase of the EU-DEMO reactor has been recently launched, with the aim of evolving the DEMO pre-conceptual layout towards a more robust and articulated geometric configuration able to cope with most of the design requirements and to show further margins for the passing of the current potential show-stoppers. Hence, the achievement of the conceptual design of the Water-Cooled Lead Lithium Breeding Blanket (WCLL BB) is one of the milestones the EUROfusion consortium aims to achieve in the close future. To this purpose, within the framework of the research activities promoted by EUROfusion, a research campaign has been launched at the University of Palermo, in close cooper…
Draining and drying process development of the Tokamak Cooling Water System of ITER
2016
Abstract The ITER Organization (IO) developed a thermal-hydraulic (TH) model of the complex first wall and blanket (FW/BLK) cooling channels to determine gas flow rate and pressure required to effectively blow out the water in the FW/BLK. In addition, US ITER conducted experiments for selected geometries of FW/BLK flow channels to predict the blowout parameters. The analysis indicates that as low as 2 MPa of pressure difference over the blanket modules will ensure substantial evacuation of the water in blankets with just a few percent remaining in the blanket flow channels. A limited validation study indicates that the analysis yields less conservative results to compare against data collec…
The DEMO water-cooled lead–lithium breeding blanket: design status at the end of the pre-conceptual design phase
2021
The Water-Cooled Lead–Lithium Breeding Blanket (WCLL BB) is one of the two blanket concept candidates to become the driver blanket of the EU-DEMO reactor. The design was enacted with a holistic approach. The influence that neutronics, thermal-hydraulics (TH), thermo-mechanics (TM) and magneto-hydro-dynamics (MHD) may have on the design were considered at the same time. This new approach allowed for the design team to create a WCLL BB layout that is able to comply with different foreseen requirements in terms of integration, tritium self-sufficiency, and TH and TM needs. In this paper, the rationale behind the design choices and the main characteristics of the WCLL BB needed for the EU-DEMO …
On the thermal dynamic behaviour of the helium-cooled DEMO fusion reactor
2019
Abstract The EU-DEMO conceptual design is being conducted among research institutions and universities from 26 countries of European Union, Switzerland and Ukraine. Its mission is to realise electricity from nuclear fusion reaction by 2050. As DEMO has been conceived to deliver net electricity to the grid, the choice of the Breeding Blanket (BB) coolant plays a pivotal role in the reactor design having a strong influence on plant operation, safety and maintenance. In particular, due to the pulsed nature of the heat source, the Primary Heat Transfer System (PHTS) becomes a very important actor of the Balance of Plant (BoP) together with the Power Conversion System (PCS). Moreover, aiming to …
On the impact of the heat transfer modelling approach on the prediction of EU-DEMO WCLL breeding blanket thermal performances
2020
Abstract The Water-Cooled Lithium-Lead Breeding Blanket is a key component of a fusion power plant, in charge of ensure Tritium production, shield Vacuum Vessel and magnets and remove the heat power deposited by radiation and particles arising from plasma. The last function is fulfilled by First Wall and Breeding Zone independent cooling systems. Several layouts of BZ coolant system have been investigated in the last years to identify a configuration that might guarantee EUROFER temperature below the limit (550 °C) and good thermal-hydraulic performances (i.e. water outlet temperature of 328 °C). A research activity is conducted to study and compare different modelling approaches to simulat…
Analysis of the Thermo-Mechanical Behaviour of the EU DEMO Water-Cooled Lithium Lead Central Outboard Blanket Segment under an Optimized Thermal Field
2022
Within the framework of the EUROfusion research activities on the DEMO Water-Cooled Lithium Lead (WCLL) Breeding Blanket (BB) design, a research study was performed to preliminarily optimize, from the thermal point of view, the WCLL Central Outboard Blanket (COB) segment in order to investigate its structural behaviour under a realistic thermal field. In particular, a study of thermal analyses was performed to optimize the Double Walled Tubes and Segment Box cooling channels’ geometric configurations along the poloidal extension of the WCLL COB segment, in order to obtain a spatial temperature distribution fulfilling the thermal design requirement. Then, the thermo-mechanical analysis of th…
The design of the DONES lithium target system
2019
Abstract In the framework of the EU fusion roadmap implementing activities, an accelerator-based Li(d,xn) neutron source called DONES (Demo-Oriented early NEutron Source) is being designed within the EUROfusion Work Package Early Neutron Source as an essential irradiation facility for testing candidate materials for DEMO reactor and future fusion power plants. DONES will employ a high speed liquid lithium jet struck by a 125 mA, 40 MeV deuteron beam to generate the intense neutron flux used to irradiate the material samples up to the desired level of displacement damage (˜10 dpa/fpy for iron in 0.3 l) and He production rates (˜10-13 appm He/dpa). In order to rapidly achieve a sound and stab…
Multiphysics Optimization for First Wall Design Enhancement in Water-Cooled Breeding Blankets
2021
Abstract The commercial feasibility of the first fusion power plant generation adopting D-T plasma is strongly dependent upon the self-sustainability in terms of tritium fuelling. Within such a kind of reactor, the component selected to house the tritium breeding reactions is the breeding blanket, which is further assigned to heat power removal and radiation shielding functions. As a consequence of both its role and position, the breeding blanket is heavily exposed to both surface and volumetric heat loads and, hence, its design requires a typical multiphysics approach, from the neutronics to the thermo-mechanics. During last years, a great deal of effort has been put in the optimization of…
A multi-physics integrated approach to breeding blanket modelling and design
2019
Abstract Often, for the design of a component, several kinds of analyses are needed. Even more frequently, the different fields of study, to be taken into account for the design verification, have to be examined minutely until the final results are satisfying. Furthermore, when geometry modifications are required, for instance to fulfill the component functions, the analyses cycle has to be restarted and an iterative process has to be carried out. This procedure may be time-consuming and herald of errors, in particular if it is demanded to the human activity. Therefore, it is more convenient for the scientific community to adopt a numerical tool that can combine various computational codes.…
Parametric study of the influence of double-walled tubes layout on the DEMO WCLL breeding blanket thermal performances
2020
Abstract Within the framework of the EUROfusion activities regarding the EU-DEMO Breeding Blanket (BB) concept, the University of Palermo is long-time involved, in close cooperation with ENEA, in the design of the Water Cooled Lithium Lead (WCLL) BB, that is one of the two concepts under consideration for the DEMO reactor. It is mainly characterized by a liquid lithium-lead eutectic alloy acting as breeder and neutron multiplier, as well as by subcooled pressurized water flowing as coolant under PWR-like conditions (pressure of 15.5 MPa and inlet/outlet temperatures of 295 °C/328 °C). A research campaign has been recently carried out to study the potential influence of the Breeding Zone coo…
Preliminary thermal-hydraulic analysis of the EU-DEMO Helium-Cooled Pebble Bed fusion reactor by using the RELAP5-3D system code
2021
Abstract In the frame of the activities promoted and encouraged by the EUROfusion Consortium aimed at developing the EU-DEMO fusion reactor, great emphasis has been placed at a very early stage of the design to incorporate the provisions needed to improve the overall plant safety and reliability performances as well as to analyse possible mitigation actions. In this framework, the research activity has been focused on the representative and safety relevant cooling loop of the Helium Cooled Pebble Bed (HCPB) Breeding Blanket (BB) Primary Heat Transfer System (PHTS), purposely selected by the safety team, in order to assess its thermal-hydraulic behaviour during normal operational conditions …
Development of a Two-Dimensional Simplified Tool for the Analysis of the Cooling of the ITER TF Winding Pack
2018
The cooling of the ITER toroidal field (TF) coils winding pack is guaranteed by the circulation of supercritical helium (He) in seven hydraulic circuits corresponding to the Nb3Sn cable in conduit conductors, and in 74 channels devoted to the cooling of the stainless steel case supporting the winding pack. A tool entirely developed inside ANSYS with the APDL language has been created with the aim of computing the temperature distribution in the TF winding pack in different poloidal locations. The tool also allows the assessment of the He temperature during plasma operation in the case cooling channels. The considered heat load is the volumetric nuclear heating computed with the MCNP code in…
Development and application of an alternative modelling approach for the thermo-mechanical analysis of a DEMO water-cooled lithium lead breeding blan…
2022
In the frame of the EUROfusion research activities devoted to the design of the DEMO breeding blanket (BB), the Water-Cooled Lithium-Lead BB (WCLL) concept is one of the candidates currently assessed in EU. To this end, an intense research campaign is ongoing to develop a robust geometric configuration for the WCLL BB Central Outboard Segment (COB). Since the current reference design of the WCLL COB segment is not mature enough to allow a full thermal-hydraulic assessment, an alternative procedure aimed at obtaining a thermal field for the whole segment without performing its complete thermal-hydraulic analysis is presented and applied in this work. The scope of the work is to obtain a ther…
Design and manufacturing feasibility of ITER TBM Frame and Dummy TBMs
2016
Abstract The operation and test of mock-ups of tritium breeding blankets relevant for a future commercial reactor is one of the goals of the ITER machine. To accomplish this goal, mock-ups of breeding blankets, called Test Blanket Modules (TBMs), are installed in three ITER equatorial ports. Each TBM and the associated shield form a TBM-set that is mechanically attached to a steel frame called TBM Frame. A Frame and two TBM-Sets form a TBM Port Plug (TBM PP). The ITER Organization is responsible for the design and manufacture of the TBM Frames and of the Dummy TBMs that could replace the TBM-sets in case they were not available. This paper describes the recent results of the design supporti…
Special Issue on Structural and Thermo-Mechanical Analyses in Nuclear Fusion Reactors
2022
Nuclear fusion is one of the most promising technologies to be adopted for the production of electricity [...]
Recent Progress in the WCLL Breeding Blanket Design for the DEMO Fusion Reactor
2018
The water-cooled lithium-lead (PbLi) breeding blanket is one of the candidate systems considered for the implementation in the European Demonstration Power Plant (DEMO) nuclear fusion reactor. This concept employs PbLi liquid metal as tritium breeder and neutron multiplier, water pressurized at 15.5 MPa as the coolant, and EUROFER as the structural material. The current design is based on the single module segment approach and follows the requirements of the DEMO-2015 baseline design. The module is constituted by a basic toroidal-radial cell that is recursively repeated along the poloidal direction where the liquid metal flows along a radial-poloidal path. The heat generated by the fusion r…
Optimization of the first wall for the DEMO water cooled lithium lead blanket
2015
The maximum heat load capacity of a DEMO First Wall (FW) of reasonable cost may impact the decision of the implementation of limiters in DEMO. An estimate of the engineering limit of the FW heat load capacity is an essential input for this decision. This paper describes the work performed to optimize the FW of the Water Cooled Lithium-Lead (WCLL) blanket concept for DEMO fusion reactor in order to increase its maximum heat load capacity. The optimization is based on the use of water at typical Pressurised Water Reactors conditions as coolant. The present WCLL FW with a waved plasma-faced surface and with circular channels was studied and the heat load limit has been predicted with FEM analy…
On the improved current pulse method for the dynamic assessment of thermal diffusive properties
2009
Fattibilità di una diversa configurazione della facility SPES-3
2013
MHD Free Convection in Helium-Cooled Lithium-Lead Blanket Modules for the Demonstration Fusion Reactor
2003
The HELICA mock-up. Experimental results and numerical thermo-mechanical analyses
2006
Out of Pile Thermo-Mechanical Testing of Breeder Pebble Beds for HCPB TBM for ITER
2005
Final report on the test campaigns HELICA I-II performed in HEFUS facility
2005
Analysis of a Generation 3+ Pressurised Water Reactor plant response to a postulated Station Black Out
2013
Thermal tests of ceramic breeder pebble beds for the Helium Cooled (HCPB) DEMO blanket
2002
On a computational study of the hydraulic behavior of the ITER divertor cassette full-scale prototype
2009
TAZZA test section. Status of the research activity at the DIN
2002
Mixed magnetohydrodynamic convection in poloidal Helium-Cooled Lithium Lead blanket modules of a fusion reactor
2004
The hydraulic behaviour of the ITER full-scale divertor cassette. The steady state analyses
2004
Water Cooled First Wall optimization for WCLL DEMO reactor
2013
Studio della risposta nucleare del Water-Cooled Lithium Lead Test Blanket Module nel reattore ITER-FEAT
2002
Studio del comportamento termico di letti di sfere tramite il metodo dell’impulso di corrente
2003
On the numerical and experimental characterisation of the ITER divertor cassette plasma facing component steady state hydraulic behaviour
2008
The hydraulic behaviour of the simulacrum of the Dome Liner of an ITER divertor cassette
2007
Hydraulic Analysis of Blanket Cooling System
2014
The hydraulic behaviour of the simulacrum of the Inner Vertical Target of an ITER divertor cassette
2007
Preparation of HEXCALIBER tests and preliminary thermo-mechanical analyses
2006
Final report on the contract EFDA 682 - Preparation of integration and hydraulic tests of full-scale divertor components.
2004
Thermo-mechanical issues induced by neutron swelling in the IFMIF Target Assembly back-plate
2011
Modifiche del Codice RELAP5 per lo Studio delle Perdite di Carico in Generatori di Vapore a Tubi Elicoidali Interessati da Flussi Bifase
2010
Analysis of the thermo-mechanical behaviour of IFMIF Target Assembly
2013
TRACE input model for SPES3 facility
2010
The hydraulic behaviour of the simulacrum of a Plasma Vessel Module of the W 7-X Reactor
2007
On the nuclear response of the HCLL-TBM in ITER-FEAT
2003
A Monte Carlo study on the possible lay-out influence in the HCLL-TBM nuclear response
2008
The hydraulic behaviour of the simulacrum of the Outer Vertical Target of an ITER divertor cassette
2007
Preliminary design of a Water Cooled Lithium Lead blanket concept for DEMO reactor
2013
On the use of an improved current pulse method for the experimental determination of the thermal diffusive properties of packed pebble beds
2005
On the computational assessment of the IFMIF-EVEDA Target Assembly thermal behaviour
2011
HEXCALIBER TEST SECTION. Preliminary thermo-mechanical analysis
2007
Hydraulic Analysis of the CDR Design of the ITER TBM Port Plug
2013
On the numerical assessment of the thermal-hydraulic behaviour of ITER Upper Port Plug Main Body cooling circuit
2010
Thermal Parametric and Thermo-Mechanical analyses for the Limiter Blanket Module of ITER
1996
On the thermal conductivity of a single size beryllium pebble bed
2002
Analisi Mediante Il Codice TRACE Delle Principali Fenomenologie Caratterizzanti Il Transitorio Conseguente Ad Una Rottura A Ghigliottina Nella Linea …
2011
ITER Upper Port Plug Draining and Drying
2010
Analysis of the thermo-mechanical behaviour of IFMIF-EVEDA Lithium Test Loop Target Assembly
2011
Piano di emergenza per il reattore nucleare di ricerca AGN-201 COSTANZA dell’Università di Palermo: presupposti tecnici e valutazioni radiologiche
2006
The hydraulic behaviour of the prototype of on a ITER divertor cassette
2009
Thermal analysis of the Helium-Cooled Lithium Lead Test Blanket Module to be irradiated in ITER
2009
Steady state and transient thermal-hydraulic analyses on ITER divertor cassette
2004
Analysis of the steady state thermal-hydraulic behaviour of the ITER blanket cooling system
2014
Modifiche del codice RELAP5/MOD3.2.b per lo studio delle perdite di carico e dello scambio termico in condotti elicoidali interessati da deflussi bif…
2011
L’attività di ricerca svolta nell’ambito della linea progettuale LP2-B.2 del programma PAR 2008-09 (CERSE III) ha visto una prima fase in cui è stato ulteriormente sviluppato il lavoro svolto nel corso dei precedenti programmi CERSE [1, 2], riguardante la validazione del codice termoidraulico avanzato Relap5/Mod3.2.b, modificato per il calcolo delle cadute di pressione in tubi elicoidali interessati da deflussi monofase e bifase, ed una seconda fase che ha comportato l’implementazione di nuove procedure valide per lo studio dello scambio termico bifase in condotti elicoidali, in aggiunta a quelle relative al solo scambio termico monofase, in precedenza implementate. Per quanto riguarda ques…
Thermo-mechanical experiment and analysis on an HCPB-TBM mock-up
2008
Thermal-hydraulic and thermal-structural analyses report of DEMO Helium-Cooled Pebble Bed blanket
2012
Education and Research in Nuclear Engineering and Radiological Protection at Nuclear Engineering Department of Palermo University
2010
The Iter Divertor Vertical Target Dummy Armor Prototype. Theoretical and Experimental Thermal-hydraulic and Thermo-mechanical Study
2001
Mixed MHD Convection and Tritium Transport in HCLL TBM Breeder Units for the ITER Fusion Reactor
2006
TAZZA mock-up pebble beds. Experimental and theoretical investigations
2002
Adeguamento dello SPES2 per Prove di Sicurezza. Analisi Preliminari per La Simulazione di un Incidente Tipo Fukushima su SPES-2
2012
Fusion pebble bed thermo-mechanical modelling. Status of the research activity at the DIN
2002
The ITER divertor cassette. Steady-state characterisation and draining and drying transient hydraulic analyses
2005
A Training Experience of Operators with the AGN-201 “Costanza” Research Reactor of Palermo University
2011
The nuclear reactor AGN-201 named “Costanza”, installed at the Nuclear Engineering Section of the Department of Energy of the University of Palermo, is a “zero power” research reactor designed to be mainly used for education purposes as well as for research applications, such as activation analysis and irradiation tests, and last, but not the least, for radionuclide production to be used in nuclear medical applications. Due to its intrinsic safety and low margin of reactivity (less than 350 p.c.m.) so as to the absence of start-up and shut-down time limits, it represents a useful training tool for operators, too. This paper reports some of the activities carried out within the framework of …
A Thermo-Mechanical Constitutive Model of the Lithium Orthosilicate and Beryllium Pebble Beds. An Application on the Thermo-Mechanical Study of the S…
2000
On the investigation of Lithium Methatitanate pebble bed thermal diffusive properties by the Improved Current Pulse method
2012
Thermomechanical analyses of the DEMO-HCPB TBM medium scale mock-up to be tested in the HE-FUS3 facility
2000
On the AGN - 201 "COSTANZA" Research Reactor at the Department of Nuclear Engineering of the University of Palermo
2005
On the use of the Porous Elasticity and the Drucker-Prager models in the numerical simulations of the pebble beds thermo-mechanical behavior
2000
Fusion pebble bed thermo-mechanical modelling. Final report A. Status of the research activity at the DIN
2002
The thermoelasticity problem of axisymmetric solids with a thermal conductivity linearly depending on the temperature and on the volumetric strain
2000
Analysis of the SPES-3 direct vessel injection line break by using TRACE code
2011
On the nuclear impact of different coolants in the HCLL-TBM in ITER-FEAT
2003
Experimental investigations on the thermal conductivity of packed pebble beds by the current pulse method
2003
Analysis of the thermo-mechanical behaviour of IFMIF target assembly integrated with its support framework
2014
On the assessment of lithium orthosilicate pebble bed thermal diffusive properties by the improved current pulse method
2008
Fusion pebble bed thermo-mechanical modelling. Final report B. Coupled thermo-mechanical analyses in presence of volumetric heat sources
2002
HEXCALIBER TEST SECTION. Thermo-mechanical analysis
2009
Thermal-hydraulic and thermal-structural analyses report of DEMO WCLL blanket
2012
Applicazione del metodo Monte Carlo a problemi monodimensionali di conduzione termica stazionaria in sistemi con conducibilità dipendente dalla posiz…
2010
The hydraulic behaviour of the ITER full-scale divertor cassette. The transient analyses
2004
Studio termomeccanico del modulo Limiter del reattore ITER-FEAT
2002
La sezione di prova HELICHETTA. Analisi termomeccaniche numeriche e sperimentali
2003
Studies on the AGN - 201 "COSTANZA" Research Reactor
2007
Sezione di prova HELICA. Analisi termomeccaniche
2005
Analysis of the thermo-mechanical behaviour of IFMIF fast disconnecting system
2013
Steady state thermo-mechanical analysis of the IFMIF Lithium target system with bayonet backplate under design conditions
2014
PROVE ED ESPERIMENTI DI VERIFICA DELLA SICUREZZA DEL REATTORE NUCLEARE DI RICERCA AGN-201 “COSTANZA”
2011
The “Zero Power” Nuclear Reactor AGN-201 “Costanza” available at Nuclear Engineering Department of Palermo University is a valuable tool for educational purpose and research. Besides being a useful tool for training of operators, without the time period limits on start-up and shut-down of the larger reactors, it allows the study of some phenomena regarding nuclear reactor physics, applied neutronics, neutron dosimetry, nuclear measurements as well as testing of nuclear instrumentation and methods. The experience of work and the obtained results highlight the simplicity of AGN-201 reactor control, its intrinsic safety and its overall versatility in various fields of Nuclear Engineering.