R. Lo Frano

Evaluation of the Fluid-Structure Interaction Effects in a Lead-Cooled Fast Reactor

Abstract The aim of the study is to investigate the structural effects induced by a beyond design basis earthquake on the main safety relevant structures and components of an isolated liquid metal reactor, such as the European Lead-cooled SYstem (ELSY) or ALFRED projects. An extensive R&D program related to heavy-metal cooled systems was recently carried out as Euratom projects of the 6th and 7th Framework Programmes, addressing many of the most important issues related to the viability of a lead-cooled fast reactor. The importance of seismic effects is mainly related to the high inertial forces of the primary coolant (liquid metal) and associated with the impact of the liquid waves on the reactor structures. The isolation devices considered for the design were represented by means of an iso-elastic approach. Moreover, the influence of isolator failure was also evaluated. The fluid-structure interaction and the sloshing phenomenon, characterized by hydrodynamic and impact forces, were numerically investigated, since an explicit analytical solution for structures of such complex geometry is not possible. Numerical calculations (i.e., dynamic nonlinear analyses) were carried out with appropriate finite element method codes and external coupling. A validation analysis was further performed to check the consistency and adequacy of the method used with respect to the American Society of Civil Engineers (ASCE) 4-98 rules. The accelerations propagated in the reactor building confirmed the favorable effect of the seismic isolation, even with 2% faulted isolators. The results indicated that the stress state, in the reactor internals, is not sufficient to impair their structural integrity, although there is localized plastic deformation.

Buckling of Imperfect Thin Cylindrical Shell under Lateral Pressure

The strength of thin shells, under external pressure, is highly dependent by the nature of imperfection. This paper investigates buckling behaviour of imperfect thin cylindrical shells with analytical, numerical, and experimental methods in conditions for which, at present, a complete theoretical analysis was not found in literature. In general, collapse is initiated by yielding, but interaction with geometrical instabilities is meaningful, in that imperfections reduce the load bearing capacity by an amount of engineering significance also when thickness is considerable. The aim of this study was to conduct experiments that are representative of buckling, in the context of NPP applications as, for instance, the IRIS (international reactor innovative and secure) and LWR steam generator (SG) tubes. At Pisa University, a research activity is being carried out on the buckling of thin walled metal specimen, with a test equipment (and the necessary data acquisition facility) as well as numerical models were set up by means FEM code. The experiments were conducted on A-316 test specimens, tubes with and without longitudinal welding. The numerical and experimental results comparison highlighted the influence of different types of imperfections on the buckling loads with a good agreement between the finite-element predictions and the experimental data.

Numerical-experimental analyses by Hot-Wire method of an alumina cylinder for future studies on thermal conductivity of the fusion breeder materials

The determination of the thermal conductivity of breeder materials is one of the main goal in order to find the best candidate material for the fusion reactor technology. Experimental tests have been and will be carried out with a dedicated experimental devices, built at the Department of Civil and Industrial Engineering of the University of Pisa. The methodological approach used in doing that is characterized by two main phases strictly interrelated each other: the first one focused on the experimental evaluation of thermal conductivity of a ceramic material, by means of hot wire method, to be subsequently used in the second phase, based on the test rig method, to determine the thermal conductivity of pebble bed material. To the purpose, two different experimental devices have been designed and built. This paper deals with the first phase of the methodology. In this framework, the equipment set up and built to perform Hot wire tests, the ceramic material (a cylinder of alumina), the experimental procedure and the measured results obtained varying the temperature, are presented and discussed. The experimental campaign has been lead from 50°C up to 400°C. The thermal conductivity of the ceramic material at different bulk temperatures has been obtained in stationary conditions (detected on the basis of the temperature values measured during the experiment). Numerical analyses have been also performed by means of FEM code Ansys©. The numerical results were in quite good agreement with the experimental one, confirming also the reliability of code in reproducing heat transfer phenomena.

Experimental thermal characterization of concrete to be used in CP5.2 packaging system

This work deals with the experimental evaluation of the thermal conductivity of a concrete matrix to be used for embedding LILW bituminised Wastes into the packaging system. Such a type of packaging, identified with the acronym CP5.2, has been also qualified by executing at the Lab. Scalbatraio of Dep. of Civil and Industrial Engineering of the University of Pisa, an experimental test campaign accordingly to the IAEA regulations. In particular, the knowledge of the thermal conductivity is important because of the fire or furnace test to carry out in oven at 800 °C for 30 minutes. These data allowed to simulate pre-test conditions and to set up safety and operational fire test procedures. The concrete thermal conductivity was obtained by performing hot wire tests on cylindrical concrete samples for temperatures ranging from 100° to about 800°C. Thermal conductivity is determined at steady state condition. Results indicate a monotonically reduction of the thermal conductivity as the temperature increases. The comparison with concrete thermal conductivity data available in literature indicates a quite good agreement. Finally, visual and X-ray inspection of sample did not highlight the presence of micro/macro damages that would have affected the thermal performance of the concrete under study.

Seismic Analysis Approach Applied to a Small Size Next Generation Nuclear Reactor

The aim of the paper is to evaluate the behaviour of a Near Term nuclear energy system example with reference to IRIS (International Reactor Innovative and Safety) project. As it is well known the development of new and future-generation nuclear power plant (Gen IV NPP) is strictly related to the sustainability, safety and reliability as well as to the proliferation resistance. In this paper, the safety aspects related to the effects of a severe earthquake (Safe Shutdown Earthquake) as well as to the induced loads are treated by means the Substructure and Time History Approaches, assuming a free field Peak Ground Acceleration equal to 0.3 g as input motion. The analyses and upgrading of the geometry structures with highest probability of criticality are performed on rather complex and detailed 3D finite element (FE) models. The main goals were: the evaluation of the dynamic characteristics of each considered structure, the verification of the load bearing structures in order to obtain a preliminary assessment of the adopted methodological approach and structural models. The analyses results and dynamic response of internal components (e.g. Nuclear Buildings, etc.) seem to confirm the possibility to upgrade the geometry and the performances of the proposed design choices.Copyright

PRELIMINARY SAFETY EVALUATION OF AN AIRCRAFT IMPACT ON A NEAR- SURFACE RADIOACTIVE WASTE REPOSITORY

The aircraft impact accident has become very significant in the design of a nuclear facilities, particularly, after the tragic September 2001 event, that raised the public concern about the potential damaging effects that the impact of a large civilian airplane could bring in safety relevant structures. The aim of this study is therefore to preliminarily evaluate the global response and the structural effects induced by the impact of a military or commercial airplane (actually considered as a “beyond design basis” event) into a near surface radioactive waste (RWs) disposal facility. The safety evaluation was carried out according to the International safety and design guidelines and in agreement with the stress tests requirements for the security track. To achieve the purpose, a lay out and a scheme of a possible near surface repository, like for example those of the El Cabril one, were taken into account. In order to preliminarily perform a reliable analysis of such a large-scale structure and to determine the structural effects induced by such a types of impulsive loads, a realistic, but still operable, numerical model with suitable materials characteristics was implemented by means of FEM codes. In the carried out structural analyses, the RWs repository was considered a “robust” target, due to its thicker walls and main constitutive materials (steel and reinforced concrete). In addition to adequately represent the dynamic response of repository under crashing, relevant physical phenomena (i.e. penetration, spalling, etc.) were simulated and analysed. The preliminary assessment of the effects induced by the dynamic/impulsive loads allowed generally to verify the residual strength capability of the repository considered. The obtained preliminary results highlighted a remarkable potential to withstand the impact of military/large commercial aircrafts, even in presence of ongoing concrete progressive failure (some penetration and spalling of the concrete wall) of the impacted area.

Evaluation of the Isolation System Influence in the Seismic Loading Analysis Applied to a Near Term Deployment Reactor

Nuclear power plant (NPP) design is strictly dependent on the seismic hazards and safety aspects related to the external events of the site. Passive vibration isolators are the most simple and reliable means to protect sensitive equipment from environmental shocks and vibrations. This paper concerns the methodological approach to treat isolation applied to a near term deployment reactor and its internals structures in order to attain a suitable decrease of response spectra at each floor along the height of the structure. The aim of this evaluation is to determine the seismic resistance capability of as-built structures systems and components in the event of the considered Safe Shutdown earthquake (SSE). The use of anti-seismic techniques, such as seismic isolation (SI) and passive energy dissipation, seems able to ensure the full integrity and operability of important structures and systems even in very severe seismic conditions. Therefore the seismic dynamic loadings, propagated up to the main reactor system and components, may be reduced by using the developed base-isolation system (high flexibility for horizontal motions) that might combine suitable dampers with the isolating components to support reactor structures and building. To investigate and analyze the effects of the mentioned earthquake on the considered reactor internals, a deterministic methodological approach, based on the evaluation of the propagation of seismic waves along the structure, was used. To the purpose of this study a numerical assessment of dynamic structural response behaviour of the structures was accomplished by means of the finite element approach and setting up, as accurately as possible, a representative three-dimensional model of mentioned NPP structures. The obtained results in terms of response spectra (carried out from both cases of isolated and not isolated seismic analyses) were compared in order to highlight the isolation technique effectiveness.Copyright

Application of PHADEC Method for the Decontamination of Radioactive Steam Piping Components

The dismantling of nuclear plants is a complex activity that originates often a large quantity of radioactive contaminated residue.In this paper the attention was focused on the PHADEC (PHosphoric Acid DEContamination) plant adopted for the clearance of Caorso NPP (in Italy) metallic systems and components contaminated by Co60 (produced by the neutron capture in the iron materials), like the main steam lines, moisture separator of the turbine buildings, etc.The PHADEC plant consists in a chemical off line treatment: the crud, deposited along the steam piping during life plant as an example, is removed by means of acid attacks in ponds coupled to a high pressure water washing.Due to the fact that the removed contaminated layers, essentially, iron oxides of various chemical composition, depend on components geometry, type of contamination and time of treatment in the PHADEC plant, it becomes of meaningful importance to suggest a procedure capable to improve the control of the PHADEC process parameters.This study aimed thus at the prediction and optimization of the mentioned treatment time in order to improve the efficiency of the plant itself and to achieve, in turn, the minimization of produced wastes. To the purpose an experimental campaign was carried out by analysing several samples, i.e. taken along the main steam piping line.Smear tests as well as metallographic analyses were carried out in order to determine respectively the radioactivity distribution and the crud composition on the inner surface of the components. Moreover the radioactivity in the crud thickness was measured. These values allowed finally to correlate the residence time in the acid attack ponds to the level of the achieved decontamination.Copyright

Seismic response of reactor vessel internals in the IRIS reactor

The intent of this paper is the presentation and discussion of a methodology for the evaluation and analysis of seismic loads effects on a nuclear power plant. To help in focussing the presented methodology, a preliminary simplified analysis of an integral, medium size next generation PWR reactor structure (IRIS project, an integral configuration PWR under study by an international group) was considered as an application example also for models/codes evaluation. The performed preliminary seismic analysis, even though by no means complete, is intended to evaluate the method of calculating the effects of dynamic loads propagation to the reactor internals for structural design as well as geometrical and functional optimisation purposes. To this goal, finite element method and separated (sub) structures approaches were employed for studying the overall dynamic behaviour of the nuclear reactor vessel. The analysis was set up by means of numerical models, implemented on the MARC FEM code, on the basis of Design Response Spectra as indicated on the relevant rules for Nuclear Power Plants (NRC 1.60) design. The seismic analysis is indented to evaluate the dynamic loads propagated from the ground through the Containment System and Vessel to the Steam Generator’s tubes.© 2006 ASME