Gianfranco Caruso

Sea-water desalination with nuclear and other energy sources: the EURODESAL project

This paper summarises our recent investigations undertaken as part of the EURODESAL project on nuclear desalination, currently being carried out by a consortium of four European, and one Canadian, industrials and two leading EU R&D organisations. Major achievements of the project, as discussed in this paper are: • Coherent demonstration of the technical feasibility of nuclear desalination through the elaboration of coupling schemes for optimum cogeneration of electricity and water and by exploring the unique capabilities of the innovative nuclear reactors and desalination technologies. • Verification that the integrated system design does not adversely affect nuclear reactor safety. • Development of codes and methods for an objective economic assessment of the competitiveness and sustainability of proposed options through comparison, in European conditions, with fossil energy based systems. Results obtained so far seem to be quite encouraging as regards the economical viability of nuclear desalination options. Thus, for example, specific desalination costs (

High-energy efficiency desalination project using a full titanium desalination unit and a solar pond as the heat supply

/m 3 of desalted water) for nuclear systems, such as the AP-600 and the French PWR-900 (reference base case), coupled to multiple effect distillation (MED) or the reverse osmosis (RO) processes, are 30–60% lower than the desalination costs for fossil energy based systems, using pulverised coal and natural gas with combined cycle, at low discount rates and recommended fossil fuel prices. Even in the most unfavourable scenarios for nuclear energy (discount rate = 10%, low fossil fuel costs) desalination costs with the nuclear reactors are 7–20% lower, depending upon the desalination capacities. Furthermore, with the advanced coupling schemes, utilising waste heat from nuclear reactors, the gains in specific desalination costs of nuclear systems are increased by another 2–15%, even without system and design optimisation. A preliminary evaluation shows that desalination costs with the GT-MHR, coupled to a MED process, could still be much lower than the above nuclear options for desalting capacities ≤ 43 000 m 3 per day. This is because its design intrinsically provides “virtually free” heat at ideal temperatures for desalination (80–100 ◦ C).

A correlation to predict chf in subcooled flow boiling

Abstract The project concerns the design, optimization, construction, assembling, start-up and extensive monitoring of an experimental plant consisting of a full-titanium desalinator coupled with a small solar pond. The operational tests took place at the site of the solar pond of the University of Ancona (Italy). Data collected during the start-up and operation of the plant under various conditions are being utilized for improving expertise on heat recovery with highly corrosive fluids, on co-generation plants aimed at producing electricity and fresh water, and on desalination fed by solar energy. Some preliminary economic evaluations are discussed.

Experimental investigation of free surface vortices and definition of gas entrainment occurrence maps

The present paper provides a discussion of the thermal-hydraulics requirements in fusion reactor components, with particular reference to the removal of high heat flux from plasma facing components and the critical heat flux (CHF) limit. Available experimental data on CHF of subcooled flow boiling in water, in the ranges of interest of fusion reactors thermal-hydraulic conditions, i.e. high inlet subcooling and velocity, and small channel diameter and length, are analyzed to discuss the influence of these parameters on CHF. The reference data-set (1887 experimental points) covers a wide range of operating conditions in the frame of present interest (0.1 < p < 8.4 MPa; 0.3 < D < 25.4 mm; 0.25 < L < 61 cm; 900 < G < 90000 kg/m2·s; 0.3 < Tin < 242.7°C). The aim of the research was to identify a new correlation based on a structure representing the relation of heat balance and using a non-linear regression analysis of the available data-set. A preliminary correlation (DINCE-92), based on 544 data points, had been developed providing a sensible improvement in predictions with respect to available predictive tools. Now, a new correlation (DINCE-93), based on the same structure of the above one and characterized by a very good statistics using a total of 1887 experimental points (88% of predictions are within ±20%) and by an R.M.S. error of 14.2%, has been identified and analyzed.

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

For the future development of Generation IV nuclear reactors, both safety and economic targets have to be achieved. In order to increase, at the same time, the power density generation and the safety features, a huge R&D effort is still required. Referring especially to Liquid Metal Cooled Fast Reactors, much attention is placed on Gas Entrainment (GE) phenomena, which could cause unlikely positive reactivity insertion accident. The GETS experimental facility (Gas Entrainment Test Section), especially aimed at studying the free surface vortices occurrence, has been built in the thermal-hydraulics laboratory of the DIAEE. The main purpose of this facility is to identify the most important parameters affecting the whirlpools formation and evolution. Experimental tests and preliminary observations have been performed. Different vortex behaviours related to different experimental conditions have been identified and presented in the present paper. 2D occurrence maps as function of different dimensionless groups (Reynolds, Froude and Weber numbers and H* = H/d ratio) have been defined. In the present paper, the results of a first experimental campaign, carried out with tap water, are discussed.

Numerical simulation of turbulent forced convection in liquid metals

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 the future generation of nuclear reactors, liquid metals are foreseen to be used as a primary coolant. Liquid metals are characterized by a very low Prandtl number due to their very high heat diffusivity. As such, they do not meet the so-called Reynolds analogy which assumes a complete similarity between the momentum and the thermal boundary layers via the use of the turbulent Prandtl number. Particularly, in the case of industrial fluid-dynamic calculations where a resolved computation near walls could be extremely time consuming and could need very large computational resources, the use of the classical wall function approach could lead to an inaccurate description of the temperature profile close to the wall. The first aim of the present study is to investigate the ability of a well- established commercial code (ANSYS FLUENT v.14) to deal with this issue, validating a suitable expression for the turbulent Prandtl number. Moreover, a thermal wall-function developed at Universite Catholique de Louvain has been implemented in FLUENT and validated, overcoming the limits of the solver to define it directly. Both the resolved and unresolved approaches have been carried out for a channel flow case and assessed against available direct numerical and large eddy simulations. A comparison between the numerically evaluated Nusselt number and the main correlations available in the literature has been also carried out. Finally, an application of the proposed methodology to a typical sub-channel case has been performed, comparing the results with literature correlations for tube banks.

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

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.

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

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.

Supercritical carbon dioxide applications: features and advantages

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.