32 research outputs found
BOOK REVIEW OF ULYSSES MOORE: DOOR TO TIME WRITTEN BY PIERDOMENICO BACCALARIO
No full textFinal Project ini merupakan ulasan dari novel Ulysses Moore: Door to Time yang telah dialih bahasakan menjadi Ulysses Moore: Pintu Waktu. Buku ini ditulis oleh seorang penulis Italia bernama Pierdomenico Baccalario. Ulysses Moore bergenre fiksi anak dan bercerita tentang petualangan sekelompok anak yang mencoba menguak misteri sebuah rumah. Door to Time merupakan seri awal dari rangkaian 9 buku Ulysses Moore. Di dalamnya terdapat sebab-sebab dan asal- muasal kejadian yang ada di buku-buku selanjutnya. Sedangkan Final Project ini sendiri menyajikan review dan pandangan-pandangan penulis mengenai buku tersebut. Ulysses Moore adalah buku anak yang sangat menantang. Gaya pembawaan buku ini menjadikan ceritanya seolah benar-benar terjadi, misal dengan adanya email dari editor di tiap pendahuluan yang berisi hasil penyelidikannya tentang Ulysses Moore. Sebaliknya, sulit membayangkan beberapa anak di awal umur belasan tahun dapat memecahkan berbagai misteri seperti yang ada di buku ini. Meski begitu, Ulysses Moore tetap layak dibaca sebagai tambahan pembendaharaan bacaan yang bagus baik untuk anak maupun dewasa
Experimental and numerical analysis of heavy liquid metal systems for Generation IV fast reactors
No full textThe present work reports the results achieved during the doctoral research activity realized in partnership between DIAEE (Dipartimento di Ingegneria Astronautica, Elettrica ed Energetica) of Sapienza University of Rome and ENEA (Agenzia nazionale per le nuove tecnologie, l’energia e lo sviluppo economico sostenibile). The activities have been carried out within the EU scientific community, since they are part of the R&D activities foreseen in the two HORIZON2020 European Projects SESAME and MYRTE.
After a brief description of the Lead Fast Reactors (LFRs) technologies and the actual status of the related R&D programs worldwide, the description of the Lead Bismuth Eutectic (LBE)-cooled pool-type facility CIRCE is presented. In particular, the work is focalized to the newest test section presently installed on CIRCE and named HERO (Heavy liquid mEtal pRessurized water cOoled tubes). The performed experimental campaigns aimed at characterizing a prototypical steam generator with double-wall bayonet tubes, evaluating its thermal-hydraulic performances in normal operational and transient scenarios.
The experimental activity on CIRCE-HERO has been supported by a numerical pre-test analysis described in the third section of this document. In particular, the RELAP5-3D© model of the HERO secondary loop has been set-up and it has been used to define the start-up procedure of the facility and to achieve feedbacks on the performances of the steam generator.
The core of this document is dedicated to the description and post-test analysis of the two experimental campaigns executed on CIRCE-HERO. The first experimental campaign, consisting of three tests, has been performed in the framework of the HORIZON2020 SESAME EU project, with the objective to support the development of the ALFRED design. The second one, consisting of nine tests, has been executed in the framework of the HORIZON2020 MYRTE EU project, with the purpose to support the development of MYRRHA and acquiring experimental data relevant for MYRRHA primary heat exchanger.
To extend the knowledge and validation of SYS-TH codes when applied for LFRs, a simulation activity has been performed in the Benchmark exercise for SYStem Thermal-Hydraulic (SYS-TH) codes and CFD/SYS-TH codes validation, in the framework of the H2020 SESAME project. A RELAP5-3D© model of the NACIE-UP facility has been set up and it has been involved to perform a preliminary blind simulation activity and a subsequent post-test analysis on the basis of the experimental results available from the test performed on NACIE-UP.
A final summary, conclusions and future perspectives are given in the final section of the document
Uncertainty quantification method for RELAP5-3D© using RAVEN and application on NACIE experiments
No full textThe best estimate plus uncertainty (BEPU) method plays a key role in the development of the innovative Generation IV nuclear reactors, for the improvement of knowledge and the good evaluation of the safety margins for new phenomena. The aim of this paper is to validate an uncertainty quantification (UQ) approach using RAVEN code. RAVEN, developed at the Idaho National Laboratory, is a multipurpose probabilistic and uncertainty quantification framework, capable to communicate with any system code, implemented with an integrated validation methodology involving several different metrics. In this activity, a coupled calculation RELAP5-3D/RAVEN has been performed to assess the validity of the embedded UQ approach. The simulations have concerned two tests conducted on NACIE (NAtural CIrculation Experiment) facility, a non-nuclear loop-type system using Lead-Bismuth Eutectic (LBE) as coolant, realized at the ENEA Brasimone Research Centre (Italy) to support the development of the GEN-IV reactors. The experimental tests are aimed to investigate the phenomena related to the natural and gas enhanced circulation flow regimes of heavy liquid metals (HLM), and to test and validate the main components in a LBE environment. A numerical 1D model of NACIE facility has been realized and the post-test analysis has been carried out using RELAP5-3D© ver. 4.3.4. Based on these results, a statistical analysis has been performed using RAVEN computer code, investigating the capability of three probabilistic comparison metrics fully integrated in the code
RELAP5-based thermal-hydraulic assessment of the STEAM facility for DEMO WCLL balance of plant analysis
No full textDEMO power station aims to demonstrate the generation of hundred MWs of electrical power from fusion reactions, then transmitted from the Tokamak reactor to the grid through the Balance of Plant (BoP). The design approach for the Water Cooled Lithium Lead (WCLL) Breeding Blanket (BB) Primary Heat Transfer Systems (PHTSs) leverages nuclear industry expertise but faces challenges due to DEMO pulsed operation and low-load periods. To assess the feasibility of these components, ENEA Experimental Engineering Division at Brasimone R.C. is designing STEAM, a facility investigating water technologies applied to the DEMO BB and BoP systems and components. STEAM is mainly composed of a primary system reproducing the DEMO WCLL BB PHTS thermodynamic conditions (15.5 MPa, 328-295 degrees C) and a secondary two-phase (liquid/steam) loop reproducing the DEMO power conversion system conditions (6.4 MPa, 238-300 degrees C). Experimental validation will reproduce steady-state and transient operation under DEMO-relevant conditions, including dedicated tests on the DEMO once through steam generator mock-up. STEAM objectives and description are presented in this paper, together with the RELAP5/Mod3.3 nodalization of the facility. The latter is used to, thermal-hydraulically characterize the facility behaviour. The outcomes of the steady-state qualification supported the optimization of the system layout appraising the performances of key components under the specified operating conditions
Il diritto di accesso ai documenti come libertà e come prerogativa
No full textTaken from a Hearing at the Chamber of Deputies’ Constitutional Committee, this article analyzes a bill designed to grant the right of access to any administrative document to any parliamentarian, by means of a minimal integration of Law No. 241 of 1990. It neglects the deep legal difference existing between the common access (governed by Law No. 241 of 1990) and the access aimed to exert the parliamentary mandate (which it tends to introduce). The first is the realization of a constitutional right to freedom; the second is the attribution of a privilege to the individual members of a representative body: thus, it represents a waive from the principle of equality that needs to be justified. The difference nature reflects the need of a discipline with opposite orientation: maximum openness for the freedom-access; maximum limitation for the privilege-access. This necessity cannot be satisfied by a minimal integration of Law No. 241 of 1990. The author considers essential to re-evaluate and possibly reorganize the proposal in the context of the instruments of parliamentary information, privileging those of collegial information. They are already sanctioned in the Rules of Procedure of the two Houses, which are also going through a rewriting process.Sviluppato da un’audizione alla Commissione affari costituzionali della Camera dei deputati, l’articolo analizza una proposta di legge volta ad attribuire a ogni parlamentare il diritto di accesso a tutti i documenti amministrativi, per minima integrazione della legge n. 241 del 1990. Essa trascura la profonda differenza di natura giuridica che separa l’accesso comune (disciplinato dalla legge n. 241 del 1990) dall’accesso funzionale all’esercizio del mandato parlamentare (che tende a introdurre). L’uno è attuazione di una libertà costituzionale; l’altro è attribuzione di una prerogativa al singolo componente dell’organo rappresentativo: deroga, dunque, al principio di eguaglianza che richiede di essere giustificata. La differenza di natura si riverbera in un’esigenza di disciplina a orientamento opposto: massima apertura per l’accesso-libertà; massima delimitazione per l’accesso-prerogativa. Tale esigenza non può essere soddisfatta attraverso un intervento di minima integrazione della legge 241 del 1990. L’autore ritiene necessario rivalutare ed eventualmente riarticolare la proposta nel contesto degli strumenti di informazione parlamentare, privilegiando quelli d’informazione collegiale. Essi sono già sanciti dai regolamenti delle due Camere, che peraltro sono in corso di riscrittur
GEN-IV LFR development: Status & perspectives
No full textSince Lead-cooled Fast Reactors (LFR) have been conceptualized in the frame of Generation IV International Forum (GIF), great interest has focused on the development and testing of new technologies related to Heavy Liquid Metal (HLM) nuclear reactors. In this frame, ENEA developed one of the larger European experimental fleet of experimental facilities aiming at investigating HLM thermal-hydraulics, coolant chemistry control, corrosion behavior for structural materials, and at developing components, instrumentations and innovative systems, supported by experiments and numerical tools. The present work aims at highlighting the capabilities and competencies developed by ENEA so far in the frame of the liquid metal technologies for GEN-IV LFR. In particular, an overview on the ongoing R&D experimental program will be depicted considering the actual fleet of facilities: CIRCE, NACIE-UP, LIFUS5, LECOR and HELENA. CIRCE (CIRColazione Eutettico) is the largest HLM pool facility presently in operation worldwide. Full scale component tests, thermal stratification studies, operational and accidental transients and integral tests for the nuclear safety and SGTR (Steam Generator Tube Rupture) events in a large pool system can be studied. NACIE-UP (NAtural CIrculation Experiment-UPgraded) is a loop with a HLM primary and pressurized water secondary side and a 250 kW power Fuel Pin Simulator working in natural and mixed convection. LIFUS5 (lithium for fusion) is a separated effect facility devoted to the HLM/Water interaction. HELENA (HEavy Liquid metal Experimental loop for advanced Nuclear applications) is a pure lead loop with a mechanical pump for high flow rates experiments. LECOR (LEad CORrosion) is a corrosion loop facility with oxygen control system installed. All the experiment actually ongoing on these facilities are described in the paper, depicting their role in the context of GEN-IV LFR development
A Note on Converse Lyapunov Theorems for Neutral Systems
No full textWe provide various extensions of recent results on the existence of Lyapunov-Krasovskii functionals for uncertain systems described by neutral functional differential equations.We consider nonlinear neutral delay systems for which the difference operator and the right-hand side of the differential equations are Lipschitz on bounded sets
Development of a Steam Generator Mock-Up for EU DEMO Fusion Reactor: Conceptual Design and Code Assessment
No full textRecent R&D activities in nuclear fusion have identified the DEMO reactor as the ITER successor, aiming at demonstrating the technical feasibility of fusion plants, along with their commercial exploitation. However, the pulsed operation of the machine causes an “unconventional” operation of the system, posing unique challenges to the functional feasibility of the steam generator, for which it is necessary to define and qualify a reference configuration for DEMO. In order to facilitate the transitions between different operational regimes, the Once Through Steam Generator (OTSG) is considered to be a suitable choice for the DEMO primary heat transfer systems, being characterized by lower thermal inertia with respect to the most common U-tube steam generators. In this framework, the ENEA has undertaken construction of the STEAM facility at Brasimone R.C., aiming at characterizing the behavior of the DEMO OTSG and related water coolant systems in steady-state and transient conditions. A dedicated OTSG mock-up has been conceived and designed, adopting a scaling procedure, keeping the height 1:1 of the DEMO OTSGs. The conceptual design has been supported by RELAP5/Mod3.3 thermal-hydraulic calculations. CFD and FEM codes have been used for fluid-dynamic analyses and mechanical stress analyses, respectively, in specific parts of the component
Status, Design and Thermal-Hydraulic Analyses of the Steam Facility for DEMO WCLL Balance of Plant Systems
No full textDEMO is a nuclear fusion power station that has, among others, the objective of demonstrating the possibility to produce several hundred MWs of electrical power from fusion reaction by the middle of this century, increasing the production of carbon-neutral electricity. In particular, the Balance of Plant of DEMO has the key role to demonstrate the feasibility of delivering the power produced within the Tokamak reactor to the grid. The design approach for Water Cooled Lithium Lead (WCLL) Breeding Blanket (BB) Primary Heat Transfer Systems (PHTSs) considers the Nuclear Power Plant experience, adopting components commonly used in nuclear industry. However, their performances could be negatively affected by the discontinuous operation (pulse-dwell-pulse) of the DEMO machine, as well as by low-load operation in dwell time. This makes mandatory a full assessment of the functional feasibility of such components throughout an accurate design and validation. ENEA Experimental Engineering Division at Brasimone R.C. supports the design and qualification of DEMO by realizing STEAM, a water operated facility conceived to investigate the water technologies applied to the DEMO BB and Balance of Plant systems and components. STEAM is mainly composed by a water primary system reproducing the DEMO WCLL BB PHTS thermodynamic conditions (15.5 MPa, 328-295°C) and a secondary two-phase (liquid/steam) water loop reproducing the DEMO power conversion system conditions (6.4 MPa, 238-300°C). The experimental validation aims at reproducing steady-state and transient operation in DEMO-relevant conditions, as well as to perform dedicated tests on a once through steam generator mock-up, representative of the one envisaged for DEMO, aiming at testing its performances during the power phases of the machine. Thermal-hydraulic analyses of STEAM have been performed by RELAP5/Mod3.3 system code. Steady-state qualification results are presented in this paper, along with the characterization of the facility dynamic behavior, realized in order to optimize the system layout and to assess the performances of the main components during the prescribed operating conditions
Alfred steam generator assessment: design and pre-test analysis of hero experiment
No full textIn the framework of the ALFRED design (Advanced Lead Fast Reactor European Demonstrator) for DEMO-LFR, a new concept of steam generator (SG) has been proposed consisting in a double wall bayonet tube bundle which improves the plant safety reducing the possibility of water-lead interaction thanks to a double physical separation between them, and allowing an easier control of eventual leakages from the coolant by pressurizing the separation region with inert gas. In order to support the development of this innovative SG configuration, the ENEA has designed and realized the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section, a mock-up (1:1 in length) which represents the ALFRED SG. This test section, implemented in the CIRCE pool facility, aims to investigate on the thermal-hydraulic features of the system, providing a database for STH codes validation. The experimental campaign consists of high pressure tests at about 180 bar carried out in the framework of the HORIZON2020 SESAME project (Simulations and Experiments for the Safety Assessment of MEtal cooled reactors). The secondary loop has been realized for the HERO SG feeding, consisting in an open loop circuit fed by demineralized water. The system is equipped with a volumetric pump and a heater in order to reach the water nominal working conditions of 335°C at the SG inlet and about 180 bar at the outlet. A preliminary test analysis is carried out by RELAP5-3D© thermal-hydraulic system code. A numerical 1-D model of the HERO SG and the secondary loop has been realized in order to test the loop layout and to characterize the main components from a thermal-hydraulic point of view, defining the start-up procedures for the achievement of the working conditions of the water for the high pressure tests. Furthermore, several simulations are carried out to investigate on the secondary system behavior both for steady states and transients
