Antonio Naviglio

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 (

Experimental data base for containment thermalhydraulic analysis

/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).

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

This paper describes the joint research project DABASCO which is supported by the European Community under a cost-shared contract and participated by nine European institutions. The main objective of the project is to provide a generic experimental data base for the development of physical models and correlations for containment thermalhydraulic analysis. The project consists of seven separate-effects experimental programs which deal with new innovative conceptual features, e.g. passive decay heat removal and spray systems. The results of the various stages of the test programs will be assessed by industrial partners in relation to their applicability to reactor conditions.

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.

The MARS nuclear reactor plant: a inherently safe, small/medium size multipurpose nuclear plant

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.

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

Abstract The crucial aspects of nuclear safety: the need of a safe shutdown and of a reliable decay heat removal system have been the starting points in the development of a medium size, inherently safe, multipurpose “new” nuclear reactor: the MARS nuclear reactor.

Safety characterstics of the MARS nuclear plant

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.

Forced convective boiling in high pressure parallel flow in tube bundles

Abstract New design criteria in the field of small and medium sized nuclear reactors will be illustrated; particular consideration will be given to the reactor MARS (Multipurpose Advanced Reactor, Inherently Safe) now under development at the Department of Energetics, University of Rome. An analysis of plant behaviour during a Station Black-out accident has been performed using the computer program RELAP 5/Mod. 2.

Development of the fast reactor MAUS for space electric generation

A comparison between the correlations usually proposed for the evaluation of boiling heat transfer coefficients outside tube bundles in high pressure conditions and experimental data (600 points) has been performed. Underpredictions of the correlations in the range of 30–40% have been found. Using the same structure as the Chen correlation but with different specifications of the Fand S corrective factors, a new correlation is proposed, whose application predicts 80% of the data with a mean error of 13%, and 90% of the data with a mean error of 30% in the following ranges: pressure, 35–90 bar; specific mass flowrate, 180–1800 kg/m2s; heat flux, 30–300 kW/m2; quality, 10–90‰.

MARS II: A design improvement to reduce construction time and cost

The results achieved through the research carried out between July 1992 and June 1994 on the ‘‘MAUS’’ project (a nuclear electric generator to be used in space missions) by the University of Rome ‘‘La Sapienza’’, are presented. In particular, a detailed description of the reactor core is reported, according to the ‘‘multi‐cell’’ fuel solution. The characteristics of fuel elements, of the reactor core, the results of the nuclear analyses carried out and the characteristic curves voltage/current of the thermionic generator are presented in the paper.