Francesco Oriolo

Statistical characteristics of a water film falling down a flat plate at different inclinations and temperatures

Abstract In this work, the statistical characteristics of the surface of a water film, freely falling down a vertical or inclined flat plate, have been investigated. The study was carried out in the frame of a research on passive cooling of heated surfaces by the evaporation of thin water films. The experiments, performed to confirm and extend previous results by the same authors, involved relatively cold water (ambient temperature or slightly warmer 20–30 °C) and warm water (50 and 70 °C). The range of Reynolds numbers includes the classical threshold for the transition between the laminar-wavy and the turbulent regimes. Two different plate inclinations with respect to the vertical position have been addressed (0° and 45°). Capacitance probes were adopted to collect discrete film thickness time series, which have been processed to extract relevant statistical data. A specific probe configuration including an electrical heating system has been developed in order to overcome the problem of vapour condensation onto the active surfaces of the electrodes in the presence of warm water. Data on mean, minimum and maximum film thickness as well as standard deviation and wave velocity are presented, discussing the trends observed as a function of film flow rate, plate inclination and film temperature, also considering the information coming from previous experimental campaigns.

Preliminary Safety Analysis of the IRIS Reactor

A deterministic analysis of the IRIS safety features has been carried out by means of the best-estimate code RELAP (ver. RELAP5 mod3.2). First, the main system components were modeled and tested separately, namely: the Reactor Pressure Vessel (RPV), the modular helical-coil Steam Generators (SG) and the Passive (natural circulation) Emergency Heat Removal System (PEHRS). Then, a preliminary set of accident transients for the whole primary and safety systems was investigated. Since the project was in a conceptual phase, the reported analyses must be considered preliminary. In fact, neither the reactor components, nor the safety systems and the reactor signal logics were completely defined at that time. Three “conventional” design basis accidents have been preliminary evaluated: a Loss Of primary Flow Accident, a Loss Of Coolant Accident and a Loss Of Feed Water accident. The results show the effectiveness of the safety systems also in LOCA conditions; the core remains covered for the required grace period. This provides the basis to move forward to the preliminary design.Copyright

Investigation of Core Degradation (COBE).

Abstract The COBE project started in February 1996 and finished at the end of January 1999. The main objective was to improve understanding of core degradation behaviour during severe accidents through the development of computer codes, the carrying out of experiments and the assessment of the computer codes’ ability to reproduce experimental behaviour. A major effort was devoted to quenching behaviour and a substantial achievement of the project was the design and commissioning of a new facility for the simulation of quenching of intact fuel rods. Two tests, carefully scaled to represent realistic reactor conditions, were carried out in this facility and the hydrogen generated during the quenching process was measured using two independent measuring systems. The codes were able to reproduce the results in the first test, where little hydrogen was generated but not the second test, where the extra steam produced during quenching caused an invigorated Zircaloy oxidation and a substantial hydrogen generation. A number of smaller parametric experiments allowed detailed models to be developed for the absorption of hydrogen and the cracking of cladding during quenching. COBE also investigated other areas concerned with late-phase phenomena. There was no experimental activity – the work included code development and the analysis of experimental data available to the project partners – either from open literature or from other projects such as Phebus-FP. Substantial improvement was made in the codes’ ability to simulate heat transfer in debris beds and molten pools and increased understanding was reached of control rod material interactions, the swelling of irradiated fuel and the movement of molten material to the lower head.

Natural Circulation of Lead-Bismuth in a One-Dimensional Loop: Experiments and Code Predictions

The paper summarizes the results obtained by an experimental and computational study jointly performed by ENEA and University of Pisa. The study is aimed at assessing the capabilities of an available thermal-hydraulic system code in simulating natural circulation in a loop in which the working fluid is the eutectic lead-bismuth alloy as in the Italian proposal for Accelerator Driven System (ADS) reactor concepts. Experiments were performed in the CHEOPE facility installed at the ENEA Brasimone Research Centre and pre- and post-test calculations were run using a version of the RELAP5/Mod.3.2, purposely modified to account for Pb-Bi liquid alloy properties and behavior. The main results obtained by the experimental tests and by the code analyses are presented in the paper providing material to discuss the present predictive capabilities of transient and steady-state behavior in liquid Pb-Bi systems.Copyright

Computational Study of Evaporative Film Cooling in a Vertical Rectangular Channel

The present work is focused on the computational analysis of evaporative film cooling, in connection with an experimental campaign carried out at the University of Pisa by the EFFE facility ( E xperiments on F alling F ilm E vaporation). The aim of the study is to contribute to the understanding of the heat and mass transfer mechanisms involved in the problem and to check the possibility of making use of a multipurpose commercial computational fluid dynamics (CFD) code for simulating mass transfer phenomena of interest in the nuclear field. After a description of the assumptions adopted in the mathematical formulation of the problem, the governing equations and the boundary conditions implemented in the code are briefly reported. In particular, the method used to evaluate the mass transfer through the interface is described. Then, the calculated results are analyzed and a comparison with experimental data is made. The improvement in the cooling capabilities of the heated plate due to evaporation with respect to the case of pure convection is also evaluated.

IRIS: Proceeding Towards the Preliminary Design

The IRIS (International Reactor Innovative and Secure) project has completed the conceptual design phase and is moving towards completion of the preliminary design, scheduled for the end of 2002. Several other papers presented in this conference provide details on major aspects of the IRIS design. The three most innovative features which uniquely characterize IRIS are, in descending order of impact: 1. Safety-by-design, which takes maximum advantage of the integral configuration to eliminate from consideration some accidents, greatly lessen the consequence of other accident scenarios and decrease their probability of occurring; 2. Optimized maintenance, where the interval between maintenance shutdowns is extended to 48 months; and 3. Long core life, of at least four years without shuffling or partial refueling. Regarding feature 1, design and analyses will be supplemented by an extensive testing campaign to verify and demonstrate the performance of the integral components, individually as well as interactive systems. Test planning is being initiated. Test results will be factored into PRA analyses under an overall risk informed regulation approach, which is planned to be used in the IRIS licensing. Pre-application activities with NRC are also scheduled to start in mid 2002. Regarding feature 2, effort is being focused on advanced online diagnostics for the integral components, first of all the steam generators, which are the most critical component; several techniques are being investigated. Finally, a four year long life core design is well underway and some of the IRIS team members are examining higher enrichment, eight to ten year life cores which could be considered for reloads.Copyright

Assessment of Scaling Principles for the Simulation of Small Break LOCA Experiments in PWRs

Abstract The choice of the scaling laws to be applied for the simulation of nuclear reactor behaviour and, more particularly, the extrapolation of data measured in experimental facilities to real plants, remains an important unresolved issue in nuclear safety. After the analysis of scaling principles adopted in the design of four PWR simulators, the above problem is dealt with in this paper. The definition of a counterpart test and a code analysis, comparing LOFT measured data with calculated trends in the PWR-PUN plant and in LOBI/MODI, LOBI/MOD2 and SEMISCALE facilities, make it possible to check the validity of the criteria utilized in the design of the experimental loops and to reduce uncertainty margins in predicting PWR behaviour.

Heat and mass transfer models in LWR containment systems

This paper summarises the basic concepts and the weaknesses of the heat and mass transfer models used in LWR containment safety analysis, with particular attention to the link between thermal-hydraulics and aerosol behaviour, a fundamental step for a realistic source term evaluation during a severe accident. A state of the art review, together with a classification of the independent variables required for an acceptable energy and mass transfer model, was initially carried out. Comparisons among the most used models, comprising semi-empirical correlations and a model based on the analogy between the momentum and the heat and mass transfer, were carried out on the basis of the experimental data available from the SOPRE II, CVTR, HDR and PHEBUS FP experimental facilities. The most significant results from this and considerations about the possibility of transferring the acquired knowledge from experimental facilities to a full-scale plant are also reported and discussed.

Experiments and Modelling Techniques for Heat and Mass Transfer in Light Water Reactors

The paper summarizes the lesson learned from theoretical and experimental activities performed at the University of Pisa, Pisa, Italy, in past decades in order to develop a general methodology of analysis of heat and mass transfer phenomena of interest for nuclear reactor applications. An overview of previously published results is proposed, highlighting the rationale at the basis of the performed work and its relevant conclusions. Experimental data from different sources provided information for model development and assessment. They include condensation experiments performed at SIET (Piacenza, Italy) on the PANTHERS prototypical PCCS module, falling film evaporation tests for simulating AP600-like outer shell spraying conditions, performed at the University of Pisa, experimental data concerning condensation on finned tubes, collected by CISE (Piacenza, Italy) in the frame of the INCON EU Project, and experimental tests performed in the CONAN experimental facility installed at the University of Pisa. The experience gained in these activities is critically reviewed and discussed to highlight the relevant obtained conclusions and the perspectives for future work.

Assessment of Transuranus fuel performance code against Studsvik Inter-Ramp BWR database

The fuel matrix and the cladding constitute the first barrier against radioactive fission product release. Therefore a defense in depth concept requires also the comprehensive understanding of fuel rod behavior and accurate prediction of the lifetime in normal operation and in accident condition as well. Investigations of fuel behavior are carried out in close connection with experimental research operation feedback and computational analyses. In this connection, OECD NEA sets up the “public domain database on nuclear fuel performance experiments for the purpose of code development and validation - International Fuel Performance Experiments (IFPE) database”, with the aim of providing a comprehensive and well-qualified database on UO2 fuel with Zr cladding for model development and code validation. This database includes the data set of the Studsvik Inter-Ramp BWR Project. The objectives of the project are to establish the failure-safe operating limits and the failure mechanism and associated phenomena, during power ramp tests, by varying the design parameters (i.e. cladding heat treatment, gap thickness and fuel density). The experimental data are used for the assessment of the Fission Gas Release (FGR) models implemented in the TRANSURANUS code versions “v1m1j07” and “v1m1j08”. The starting point of the activity is the availability of a “new” transient fission gas release model, the “TFGR model”, specifically implemented in the last code version, to cover power ramp conditions. The paper presents the complete set of simulations of all twenty rods irradiated in the R2 research reactor and the corresponding comparisons with the experimental data. Sensitivity calculations are also performed to address the influence of geometric parameters and the choice of the different code options, relevant to model the FGR, on results.© 2009 ASME