Fabio Giannetti

New RELAP5-3D Lead and LBE Thermophysical Properties Implementation for Safety Analysis of Gen IV Reactors

The latest versions of RELAP5-3D© code allow the simulation of thermodynamic system, using different type of working fluids, that is, liquid metals, molten salt, diathermic oil, and so forth, thanks to the ATHENA code integration. The RELAP5-3D© water thermophysical properties are largely verified and validated; however there are not so many experiments to generate the liquid metals ones in particular for the Lead and the Lead Bismuth Eutectic. Recently, new and more accurate experimental data are available for liquid metals. The comparison between these state-of-the-art data and the RELAP5-3D© default thermophysical properties shows some discrepancy; therefore a tool for the generation of new properties binary files has been developed. All the available data came from experiments performed at atmospheric pressure. Therefore, to extend the pressure domain below and above this pressure, the tool fits a semiempirical model (soft sphere model with inverse-power-law potential), specific for the liquid metals. New binary files of thermophysical properties, with a detailed mesh grid of point to reduce the code mass error (especially for the Lead), were generated with this tool. Finally, calculations using a simple natural circulation loop were performed to understand the differences between the default and the new properties.

Pool temperature stratification analysis in CIRCE-ICE facility with RELAP5-3D© model and comparison with experimental tests

In the frame of heavy liquid metal (HLM) technology development, CIRCE pool facility at ENEA/Brasimone Research Center was updated by installing ICE (Integral Circulation Experiments) test section which simulates the thermal behavior of a primary system in a HLM cooled pool reactor. The experimental campaign led to the characterization of mixed convection and thermal stratification in a HLM pool in safety relevant conditions and to the distribution of experimental data for the validation of CFD and system codes. For this purpose, several thermocouples were installed into the pool using 4 vertical supports in different circumferential position for a total of 119 thermocouples [1][2].The aim of this work is to investigate the capability of the system code RELAP5-3D© to simulate mixed convection and thermal stratification phenomena in a HLM pool in steady state conditions by comparing code results with experimental data. The pool has been simulated by a 3D component divided into 1728 volumes, 119 of which are centered in the exact position of the thermocouples. Three dimensional model of the pool is completed with a mono-dimensional nodalization of the primary main flow path. The results obtained by code simulations are compared with a steady state condition carried out in the experimental campaign. Results of axial, radial and azimuthal temperature profile into the pool are in agreement with the available experimental data Furthermore the code is able to well simulate operating conditions into the main flow path of the test section.

An Experimental Study on the Air-Side Heat Transfer Coefficient and the Thermal Contact Conductance in Finned Tubes

The objective of this experimental study is to evaluate the heat transfer coefficient outside a tube with annular transverse fins, derived from strips of copper mechanically bound and coupled outside. Water is used as the heating medium, in turbulent conditions and flowing at different temperatures inside the tube. Petukhovs correlation has been selected to calculate the water heat transfer coefficient in the tube. The experimental data obtained are compared with a correlation from literature, and a similar trend is observed. A fitting of the data provides a correlation for the three tubes of different external diameter (30 mm, 22 mm, and 15.6 mm) that agrees very well with the experimental values. The thermal contact conductance is identified as the main reason for the difference between data and the original Briggs and Young correlation. An estimation of the contact conductance between fins and tubes provides values between 3500 and 11000 W/m2-K, slightly increasing with the air Reynolds number (based on the external diameter of the tube), whose range is 2000 to 8000. The thermal contact resistance is estimated and its importance is confirmed, contributing 30 to 50% to the total air-side thermal resistance in the tubes used in the experiments.

Pre-test analysis of protected loss of primary pump transients in CIRCE-HERO facility

In the frame of LEADER project (Lead-cooled European Advanced Demonstration Reactor), a new configuration of the steam generator for ALFRED (Advanced Lead Fast Reactor European Demonstrator) was proposed. The new concept is a super-heated steam generator, double wall bayonet tube type with leakage monitoring [1]. In order to support the new steam generator concept, in the framework of Horizon 2020 SESAME project (thermal hydraulics Simulations and Experiments for the Safety Assessment of MEtal cooled reactors), the ENEA CIRCE pool facility will be refurbished to host the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section to investigate a bundle of seven full scale bayonet tubes in ALFRED-like thermal hydraulics conditions. The aim of this work is to verify thermo-fluid dynamic performance of HERO during the transition from nominal to natural circulation condition. The simulations have been performed with RELAP5-3D© by using the validated geometrical model of the previous CIRCE-ICE test section [2], in which the preceding heat exchanger has been replaced by the new bayonet bundle model. Several calculations have been carried out to identify thermal hydraulics performance in different steady state conditions. The previous calculations represent the starting points of transient tests aimed at investigating the operation in natural circulation. The transient tests consist of the protected loss of primary pump, obtained by reducing feed-water mass flow to simulate the activation of DHR (Decay Heat Removal) system, and of the loss of DHR function in hot conditions, where feed-water mass flow rate is absent. According to simulations, in nominal conditions, HERO bayonet bundle offers excellent thermal hydraulic behavior and, moreover, it allows the operation in natural circulation.

Analysis of unmitigated large break loss of coolant accidents using MELCOR code

In the framework of severe accident research activity developed by ENEA, a MELCOR nodalization of a generic Pressurized Water Reactor of 900 MWe has been developed. The aim of this paper is to present the analysis of MELCOR code calculations concerning two independent unmitigated large break loss of coolant accident transients, occurring in the cited type of reactor. In particular, the analysis and comparison between the transients initiated by an unmitigated double-ended cold leg rupture and an unmitigated double-ended hot leg rupture in the loop 1 of the primary cooling system is presented herein. This activity has been performed focusing specifically on the in-vessel phenomenology that characterizes this kind of accidents. The analysis of the thermal-hydraulic transient phenomena and the core degradation phenomena is therefore here presented. The analysis of the calculated data shows the capability of the code to reproduce the phenomena typical of these transients and permits their phenomenological study. A first sequence of main events is here presented and shows that the cold leg break transient results faster than the hot leg break transient because of the position of the break. Further analyses are in progress to quantitatively assess the results of the code nodalization for accident management strategy definition and fission product source term evaluation.

Sizing of the Vacuum Vessel Pressure Suppression System of a Fusion Reactor Based on a Water-Cooled Blanket, for the Purpose of the Preconceptual Design

A methodology to preliminarily evaluate the size of the suppression tank and the relief pipes for a Vacuum Vessel Pressure Suppression System, to be adopted in a fusion reactor based on a water cooled blanket, is presented. The volume of the ST depends on the total energy of the water cooling system and it can be sized based on a required final pressure at equilibrium, by a simple energy balance. The pressure peak in the VV depends mainly on break area and the flow area of the relief pipes and some suggestions about the method for a preliminarily evaluation of their size are discussed. The computer code CONSEN has been used to perform a parametric study and to verify the methodology.

Scenario Analysis on the Benefits of Multi-National Cooperation for the Development of a Common Nuclear Energy System Based on PWR and LFR Fleets

Financial aspects, environmental concerns and non-favorable public opinion are strongly conditioning the deployment of new Nuclear Energy Systems across Europe. Nevertheless, new possibilities are emerging to render competitive electricity from Nuclear Power Plants (NPPs) owing to two factors: the first one, which is the fast growth of High Voltage lines interconnecting the European countries’ national electrical grids, this process being triggered by huge increase of the installed intermittent renewable electricity sources (Wind and PV); and the second one, determined by the carbon-free constraints imposed on the base load electricity generation. The countries that due to public opinion pressure can’t build new NPPs on their territory may find it profitable to produce base load nuclear electricity abroad, even at long distances, in order to comply with the European dispositions on the limitation of the CO2 emissions. In this study the benefits from operating at multinational level with the deployment of a fleet of PWRs and subsequently, at a proper time, the one of Lead Fast Reactors (LFRs) are analyzed. The analysis performed involves Italy (a country with a current moratorium on nuclear power on spite that its biggest utility operates NPPs abroad), and the countries from South East and Central East Europe potentially looking for introduction or expansion of their nuclear power programmes. According to the predicted evolution of their Gross Domestic Product (GDP) a forecast of the electricity consumption evolution for the present century is derived with the assumption that a certain fraction of it will be covered by nuclear electricity. In this context, evaluated are material balances for the front and the back end of nuclear fuel cycle associated with the installed nuclear capacity. A key element of the analysis is the particular type of LFR assumed in the scenario, characterized by having a fuel cycle where only fission products and the reprocessing losses are sent for disposition and natural or depleted uranium is added to fuel in each reprocessing cycle. Such LFR could be referred to as “adiabatic reactor”. Owing to introduction of such reactors a substantive reduction in uranium consumption and final disposal requirements can be achieved. Finally, the impacts of the LFR and the economy of scale in nuclear fuel cycle on the Levelized Cost of Electricity (LCOE) are being evaluated, for scaling up from a national to a multinational dimension, illustrating the benefits potentially achievable through cooperation among countries.Copyright