RELAP5 simulation of two-phase flow experiments in vertical helical tubes

In the framework of the studies concerning the thermalfluid dynamic phenomena in helicoidal pipes of the innovative nuclear reactor IRIS steam generators, the Department of Nuclear Engineering of the University of Palermo in collaboration with the Politecnico di Torino Department of Energetics has been engaged in a work aimed to adapt, by implementing new suitable models, RELAP5/mod3.2.2β code to simulate the thermalfluid-dynamics and geometries such as the ones involved in helicoidal pipes. In fact this code is based on one-dimensional thermal-hydraulic relationships and presents limitations to model complicated geometry such as helicoidal pipes. Therefore the code was improved with additional correlations that are valid for two-phase flow and allow to overcome the drawbacks. The validation work of the models that were added is based on the experimental data carried out at the Politecnico di Torino Department of Energetics. In this paper it will be shown that the so modified RELAP5 code allows to represent adequately the experimental data

Dry deposition models for radionuclides dispersed in air: a new approach for deposition velocity evaluation schema

In the framework of a National Research Program funded by the Italian Minister of Economic Development, the Department of Energy, Information Engineering and Mathematical Models (DEIM) of Palermo University and ENEA Research Centre of Bologna, Italy is performing several research activities to study physical models and mathematical approaches aimed at investigating dry deposition mechanisms of radioactive pollutants. On the basis of such studies, a new approach to evaluate the dry deposition velocity for particles is proposed. Comparisons with literature experimental data show that the proposed dry deposition scheme allows to capture the main phenomena involved in the dry deposition process successfull

Un nuovo approccio per l’integrazione dell’analisi di rischio nella valutazione degli impatti ambientali connessi alle operazioni di smantellamento di un impianto nucleare

È ampiamente noto che il quadro normativo di riferimento internazionale sulle procedure autorizzative per i progetti di decommissioning degli impianti nucleari, a seguito della loro definitiva chiusura, vincola la predisposizione da parte degli enti competenti di piani che non possono prescindere dalla garanzia della sicurezza della popolazione e dei lavoratori coinvolti nelle varie procedure operative e, più in generale, della sostenibilità ambientale. Tuttavia, non esiste allo stato attuale una metodologia capace di classificare la rilevanza dei vari tipi di impatto, soprattutto dal punto di vista della analisi di rischio, ma piuttosto una varietà di strategie che dipendono fortemente dalla normativa vigente nel paese di appartenenza. L’operazione di decommissioning di un impianto nucleare è un processo lungo e complesso, che presuppone competenze multidisciplinari specialistiche e tecnologie innovative. Le attività sono piuttosto diverse da quelle svolte per la gestione dell’impianto durante la sua fase operativa, per cui gli aspetti non routinari, o comunque non ampiamente convalidati, fanno sì che il rischio di eventi indesiderati, a causa ad esempio di errori umani, risulti essere molto elevato. La pianificazione delle varie operazioni non può, quindi, prescindere dall’uso di un approccio sistematico capace di integrare l’analisi degli impatti ambientali con l’analisi del rischio, soprattutto se si ci trova ad affrontare potenziali situazioni incidentali o condizioni pericolose. I rischi possono essere oggetto di una risposta in termini di programmazione di procedure e misure di sicurezza solo se proattivamente individuati. In questo ambito, viene proposto un approccio metodologico, denominato Environmental Impact Mode and Criticality Analysis (EIMCA), che integra l’analisi degli impatti ambientali con l’analisi del rischio. Tale approccio supporta l’analista nello svolgimento di una Valutazione di Impatto Ambientale, attraverso lo studio dei possibili scenari che, durante le varie fasi programmate per lo smantellamento dell’impianto, possono indurre effetti negativi sui fattori ambientali e, quindi, situazioni pericolose per la salute degli operatori e della popolazione

Analysis of protected accidental transients in the EFIT reactor with the RELAP5 thermal-hydraulic code

The European Facility for Industrial Transmutation (EFIT) is aimed at demonstrating the feasibility of transmutation process through the Accelerator Driven System (ADS) route on an industrial scale. The conceptual design of this reactor of about 400 MW thermal power is under development in the frame of theEuropean EUROTPANS Integrated Project of the EURATOM Sixth Framework Program (FP6). EFIT is a pool-type reactor cooled by forced circulation of lead in the primary system where the heat is removed by steam generators installed inside thereactor vessel. The reactor power is sustained by a spallation neutron source supplied by a proton beam impinging on a lead target at the core centre. A safety-related Decay Heat Removal (DHR) system provided with four independent inherently safe loops is installed in the primary vessel to remove the decay heat in case of loss of secondary circuits heat removal capability. A quite detailed model of the EFIT reactor has been developed for the RELAP5 thermal-hydraulic code to be used in preliminary accidental transient analyses aimed at verifying the validity of the adopted solutions for the current reactor design with respect to the safety requirements, and confirm the inherent safety behavior of the reactor, such as decay heat removal in accidental conditions relying on natural circulation in the primary system. The accident analyses for the EFIT reactor include both protected and unprotected transients, on whether the reactor automatic trip, consisting in proton beam switch off, is actuated or not by the protection system. In this paper, the main results of the analyses of some protected transients with RELAP5 are presented. The analyzed transients concern the Protected Loss of Heat Sink (PLOHS), in which the DHR system plays a key role in bringing the reactor insafe conditions, and the Protected Loss of Flow (PLOF) transients with partial or total loss of forced circulation in the primary system