1,721,130 research outputs found
HE-FUS3 Loop: Data Sheets for Code Modelling and Experimental Data for Code Assessment
No full textThis report contains a detailed descriptions of the HE-FUS 3 helium loop and of the experimental tests performed to assess its performances in steady state and transient conditions. Moreover, all the experimental data acquired during the tests are enclosed to this report. The report, which is issued within the framework of the SP Safety (WP1) 01 RAPHAEL for the organization of a benchmark exercise aimed at the validation of the system transient analysis codes lar VHTRs, has the purpose lo provide the benchmark participants will the needed information for the HE-FUS 3 loop modelling, as well as to offer the data for code-to-experiment comparisons
Assessment of RELAP5 Mod3.3 on HE-FUS3 LOFA transients
No full textIl documento è stato redatto in collaborazione da ENEA ed ANSALDO nell'ambito del progetto europeo RAPHAEL e riguarda la partecipazione comune ad un esercizio benchmark sui codici di sistema per l' analisi in regime transitorio di Reattori a Gas ad Alta Temperatura (HTGR) utilizzando i dati sperimentali dell'impianto HE_FUS3 (ENEA Brasimone). Il codice utilizzato nel benchmark è stato RELAP5, il documento contiene una descrizione delle attività intraprese per lo sviluppo del modello dell'impianto e dei risultati ottenuti ella simulazione dei due transitori LOFA (Loss of Flow Accident) oggetto del benchmark
Set up and Preliminary Assessment of a 3D Numerical Model for the Thermo-Fluid Dynamics Analysis of an Open Square Lattice Core of a Lead Cooled Reactor
Full text linkThis report describes the set up of a simplified three-dimension numerical model for the thermo-fluid dynamic analysis of an open square lattice reactor core lead cooled on the basis of a three-dimension CFD computer code developed by DIENCA UniBo. A preliminary assessment of the model has been performed by comparing its results with the T/H reactor core behavior predicted with a one-dimension independent channel model based on RELAP5 code. To this purpose the conceptual design of the LFR core developed in the framework of the ELSY EU collaborative project has been adopted as a reference. Moreover, in order to support the set-up of the model, some results of the CFD analysis for fuel bundle of liquid metal reactors have been considered
Critical review of heat transfer models for LM in TH system codes
No full textThis document aims Io review the studies mainly conducted in previous European programs on the heat transfer models to be used in T/H system codes far the safety assessment of Liquid Metal Reactors. An overview af a wide set of empirical correlations for single-phase heat transfer in tubes and rod bundies for heavy Iiquid metal is given. The results of the previous studies performed to evaluate these correlations on the basis of the experimental data available in literature and with the support of CFD analyses are summarized. In addition, the experimental data coming from NACIE test facility, located at the ENEA Brasimone research center, are used for a comparative assessment af correiatians both in gas enhanced and natural circulation conditians. Due Io the prototypical characteristics of the NACIE test section, a rigorous analysis cannot be carried out. Anyway, the results have shown that the same carrelations that best fit the experimental results in forced convection can be used with the same confidence even in natural circulation conditions
HE-FUS3 Experimental Campaign far the Assessment of Thermal-Hydraulic Codes: Pre-Test Analysis and Test Specifications.
Full text linkThis report deals with the design of an experimental campaign to be conducted in the HE-FUS3 loop (CR BRASIMONE) in order to provide an experimental data base for the assessment of thermal-hydraulic codes used for HTR and VHTR design and safety analysis. In order to support the definition of the test matrix a pre-test activity has been carried out with the T/H system code RELAP5. To this aim a RELAP5 model of the loop (the related input deck is provided in attachment A of this report) has been developed taking advantage of the results ol previous assessment activities and already available experimental data. The pre-test activity has allowed defining a set of transients representative of operational and accident conditions that are of particular meaning for the assessment of T/H codes: the plant start-up by steps, 2 Loss of Flow Accidents with a different dynamic of the accidental event, 2 Loss of Coolant Accidents at different loop pressure. On the basis of the calculation results reported in the chapter 4 it has been possible to draw up the test specifications in attachment B taking into account the requirement for the operation of the HE-FUS3 loop and the actual conditions of the instrumentation implemented in the facility
Sviluppo e validazione del codice di calcolo per la termoidraulica di sistema CATHARE-2 per reattori refrigerati a metallo liquido pesante
Full text linkNel corso della seconda annualità dell'Accordo di Programma tra ENEA e il Ministero dello Sviluppo Economico, il codice CATHARE venne modificato con l'aggiunta delle proprietà termofisiche dei metalli liquidi piombo e lega eutettica piombo-bismuto e venne fornita una prima validazione su dati sperimentali provenienti da una facility coreana (HELIOS). Il presente documento, prodotto nel corso della terza annualità (Linea Progettuale LP3 obiettivo E1), rappresenta la naturale prosecuzione dell'attività di validazione e intende dare un quadro aggiornato sulle potenzialità di simulazione dei sistemi raffreddati a metallo liquido pesante del codice termoidraulico di sistema CATHARE. Saranno illustrate le soluzioni a problematiche evidenziate in precedenza e ne saranno avanzate nuove, inoltre i progressi ottenuti saranno testati attraverso il confronto con dati sperimentali della facility NACIE (Brasimone) e HELIOS (S.Korea)
HE-FUS3 benchmark specifications
No full textThe present document provides a detailed description of the HE-FUS3 loop - an helium cooled facility located at Brasimone Research Center in Italy - in order to support the benchmark participants in their own codes loop modelling. The experimental data make available for the benchmark are: seven steady-states at different conditions, useful to assess the thermalhydraulic models, and the initial and boundary conditions of a LOFA through the compressor slow-down and LOFA through bypass valve opening for transient simulation
Implementation and validation of the NURISP platform
Full text linkLa piattaforma NURISP e un progetto europeo sviluppato per simulare il comportamento termoidraulico dei componenti degli impianti nucleari e facility sperimentali. Questa piattaforma è utilizzata per analizzare situazioni fisiche a diverse scale e consiste di numerosi codici che vengono utilizzati per generare mesh, visualizzare i risultati e risolvere equazioni che descrivono lo stato del sistema da simulare. La piattaforma di gestione SALOME e i codici TRIO_U, NEPTUNE e CATHARE sono stati installati sul computer cluster CRESCO ENEA-GRID che si trova a Portici. TRIO_U e il codice NEPTUNE risolvono le equazioni termoidrauliche in geometrie multidimensionali con particolare attenzione al flusso bifase. Il codice CATHARE è un codice a parametri concentrati che viene utilizzato per studiare il comportamento dipendente dal tempo di sistemi complessi. In questo rapporto sono discussi l'implementazione e la validazione della piattaforma effettuata attraverso il confronto con dati sperimentali. Il codice NEPTUNE è validato con dati sperimentali dell'impianto PERSEO e il codice TRIO_U con alcuni dati sperimentali riguardanti una bolla che si stacca da pareti riscaldate. CATHARE è validato su dati sperimentali dell'impianto SPES-99
Implementation of Thermo-Physical Properties and Thermal-Hydraulic Characteristics of Lead-Bismuth Eutectic and Lead on CATHARE Code
Full text linkThe nuclear innovative systems cooled by Lead-Bismuth Eutectic (LBE) and pure Lead are object of an ongoing interest in Europe, but also outside, evidenced by a large number of national and international projects. Within the European Framework Programmes, hence, it has been highlighted the needs of a thermal-hydraulic system code able to treat these Heavy Liquid Metals (HLM) systems, with a particular interest for a 'European' code. Taking into account this scenario a Specific Topic of Cooperation (STC) has been agreed in the frame of ENEA/CEA collaboration on Nuclear Fission with the objective of extend the capability of the French system code CATHARE to simulate HLM reactors. The CATHARE code has been already modified in the recent past to be multi-fluid with well proven capabilities and moreover, it is already part of an European simulation platform (NURISP project) devoted to LWR system studies and aiming to be extended also to advanced reactors. For what regards the computational phase, the implementation takes advantage of the work already done by the Cathare-Team with the properties of sodium and for what regards the material properties, thanks to the European working group WG-LBE that in the past years have collected the state-of-the-art for LBE and Lead characteristics. This technical report summarize the work done to implement the LBE and Lead properties in CATHARE code, in collaboration/supervision with the Cathare-Team, pointing out the constraints and needs of the computational phase, the physical properties adopted, the subroutine changed and finally, some preliminary results on analytical tests
Assessment of the RELAP5 Mod3.3 capability to simulate thermal-hydraulic and dynamic behavior of helium cooled loops
No full textThis report describes the assessment of the RELAP5 code capability for HTR and VHTR design and safety analysis performed within the framework of the ENEA/MSE research program. To this purpose some tests in steady state and transient conditions were conducted in the HE-FUS3 loop (CR BRASIMONE) to provide an experimental data base characterizing the loop thermal-hydraulics and its behavior in safety relevant transient conditions. The RELAP5 model of the loop was developed taking advantage of the results of previous assessment activities and already available experimental data. By means of a pre-test analysis a set of transients representative of operational and accident conditions was defined: the plant start-up by steps, 2 Loss of Flow Accidents with a different dynamic of the accidental event, 2 Loss of Coolant Accidents at different loop pressure. The post-test analysis was carried out to assess that the RELAPS code is able to correctly reproduce the gas system thermal-hydraulics and dynamics, as well as to provide recommendations on gas cooled system modelling for the development of consistent numerical models for the HTR and VHTR accident analysis. Moreover, highlighting the weaknesses present in some aspects of the simulations performed it has been possible to suggest the needs of the future code developments
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