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Most of today's operating nuclear plants that were originally designed for 30 or 40-year life are facing the long-term operation issues. Therefore, it is of meaningful importance to assess the time-dependent degradation and the aging of the relevant nuclear systems, structures, and components because of resulting loss of structural capacity. In this framework, the inverse method is implemented starting from temperatures at an accessible boundary, which are measured through a monitoring system. The reconstruction technique uses the elaborated signal provided by the monitoring system to determine temperature at inaccessible surface: this is the so-called inverse heat transfer problem. The inverse space marching method is applied. Analytical and numerical studies are performed taking into account thermal transient conditions in order to determine thermal loads. In particular, the developed code demonstrates to be able to reconstruct temperature and stress profiles in any section of the pipe with a good accuracy. In addition, the thermal loads obtained suggest that the investigated transient condition is not able to jeopardize the integrity of nuclear power plant, confirming the possibility of the plant extension of life.

The Long-Term Operation (LTO) process involves a full screening of structures, systems, and components (SSCs), for ageing assessment in order to verify their residual safety margin is still acceptable. At today, 46% of the operating Nuclear Power Plant (NPP) has lifetime between 31 and 40 years, while 19% is in operation since more than 40 years. LTO currently represents the highest priority for all the nuclear industry. To propose and plan suitable management strategies, first step is to intensify the efforts for studying phenomena that influence the performance of SSCs and, in turn, may threat the plant safe operation. This study deals with the investigation of the performance of a primary piping (Class 1 component of a 2nd Generation PWR) subjected to the effects of alteration of material properties as caused by ageing. In this study it is proposed a numerical investigation of a piping characterized by a complex geometry. Numerical analyses were performed by means of MSC©MARC FE code. The (quantitative) influence of ageing and corrosion processes onto bent pipe are so studied. Particularly, corrosion effects generated from an operation of beyond 35 years are considered. The methodology and results may have an influence on future issues about LTO of NPPs.

The production of reliable, stable, and durable breeder blanket material is one of the most important open challenge of fusion reactor technology. A new low temperature fabrication method for producing lithium orthosilicate (Li4SiO4) pebbles based on the drip casting was developed at DICI - University of Pisa. The pebbles were produced starting from a suspension of Li4SiO4 precursors prepared by sol-gel method, which was dripped through the experimental device and coagulated into solid spheres in a saline solution. Li4SiO4 pebbles were finally obtained after sintering at high temperature. The pebbles produced are stable and well-sized (about 1.5 mm diameter). Pebbles have been analyzed in order to determine their physical and mechanical characterization. To the aim, SEM and X-ray diffractometer analyses were performed as well as compression tests by Instron machine. The pebbles obtained from the drip casting have a mechanical strength (about 11 N) comparable with that of the others solid breeder in pebbles form. In addition, if compared to these latter they have same roundness and sphericity. Results confirmed the feasibility of drip casting forming technique, and provided further useful information for the optimization of the proposed new fabrication process.

The Direct Contact Condensation (DCC) is the main phenomenon characterizing the steam condensation. It plays an important role for the operation of Vacuum Vessel Pressure Suppression System (VVPSS) tanks, particularly for managing the Ingress of Coolant Event (determining fusion reactor overpressurization). It is safety relevant (key) system of the fusion reactor because by condensing the steam generated during such accident event allows to damp the overpressure. This paper deals with experimental and theoretical analyses of the DCC at sub-atmospheric pressure. The similitude analysis was elaborated to scale up the experimental results obtained in the reduced scale facility: similitude laws are used for the design of large experimental rig. Correlations are defined starting from the water temperature and pressure variation already obtained in the small-scale rig. Furthermore, the experimental rig and its main components accordingly designed (and under construction at the University of Pisa) allow to study at large scale the steam condensation. The testing conditions are presented and discussed.

The paper studies influence the ageing effects on the failure of a Reactor Pressure Vessel (RPV) during a severe accident with a core meltdown in a Nuclear Power Plant (NPP). The studied plant is a generic high-power Generation III Pressurized Water Reactor (PWR) developed in the frame of the EU NARSIS project. A Total Station Blackout (SBO) accident was simulated with MELCOR 2.2 severe accident integral computer code. Results of the analysis, temperatures in the lower head and pressures in the lower plenum were used as initial and boundary conditions for the Finite Element Method (FEM) simulations. Two FEM models were developed, a simple two-dimensional axis-symmetric model of the lower head to study fundamental phenomena and complex 3D model to include interactions with the RPV and reactor internals. Ageing effects of a lower head were incorporated into the FEM models to investigate its influence onto lower head response. The ageing phenomena are modelled in terms of degraded mechanical material properties as σ(T), E(T). The primary outcome of the study is the quantitative estimation of the influence of ageing process onto the timing of reactor vessel failure. Presented novel methodology and results can have an impact on future consideration about Long-Term Operation (LTO) of NPPs.

For some specific applications, the pressure suppression systems should be maintained at sub-atmospheric pressure, such as in fusion plants systems. This operation conditions differ considerably from those experienced in the suppression pools of BWR operating at atmospheric pressure. The steam condensation in water at sub-atmospheric conditions is not sufficiently known, if not at all; therefore its effectiveness has to be investigated experimentally. In fact the steam condensation at sub-atmospheric pressure requires a subcooling capacity 2.5 times higher than that at atmospheric pressure. In consideration of that, a low scale experimental facility was designed and built at the University of Pisa. More than 350 steam condensation tests were performed in broad influencing parameters range (e. g. pressure within the condensation tank, water temperature and steam mass flux) to study of effectiveness of the steam condensation. In doing that, different sparger configurations (with single or multiple holes) were considered to check possible interference of condensing steam jets and bubble coalescence. Results indicate a high efficiency of condensation for all the examined conditions. The 3D map of the condensation regimes demonstrate they are dependent only on the downstream exit pressure, the water temperature and the steam mass flux per hole.

In the past decades many R&D efforts have been spent in the development of a suitable Li4SiO4 fabrication method (e.g., melt spray process, graphite bed method, capillary-based microfluidic wet method etc.), nevertheless we are still far from an “industrial standard solution”. The aim of the paper is to develop a new fabrication method capable to produce stable and well-sized pebbles of lithium orthosilicate (Li4SiO4) based on the drip casting. This method is mainly based on the dripping at room temperature, which is novel in the framework of available fabrication processes requiring high temperature: this latter is demonstrated to affect the final product characteristics. It is worthy to remark that the Li4SiO4 is a candidate material for the breeding blanket material of the fusion power reactor. In the paper we will describe the experimental apparatus, designed and built at DICI - University of Pisa with the collaboration of Bitossi Industries, and the procedure adopted in order to produce pebbles of Li4SiO4.

The condensation of steam in water may cause pressure oscillations, vibrations, and damages to piping and, in some extreme conditions to the internals of suppression tank. Significant R&D activity, especially focusing on the emergency cooling system in BWRs, has been conducted in the past decades to clarify the mechanism of the condensation oscillation evolving/related to the Direct Contact Condensation (DCC). The present paper deals with the pressure oscillations due to the condensation of steam jet in water, at sub-atmospheric condition; phenomena that have not been fully investigated yet. Vibrations, arisen as flow patterns transformed from stable to unstable, are investigated with particular attention to the dynamic behaviour of the whole suppression system, behavior of the interface, flow patterns etc. To the purpose, a numerical study is performed by means of suitable FEM in order to identify the dominant frequency associated to the steam jet condensation oscillation in water flow, and determine the resulting values of pressure and acceleration. The obtained results allowed to formulate a correlation between the dominant frequency and the condensation driving potential and steam mass flux.

The International Thermonuclear Experimental Reactor (ITER) Vacuum Vessel Pressure Suppression System (VVPSS) limits the Vacuum Vessel (VV) internal pressure, in case of loss of coolant (LOCA) or other pressurizing accidents from the in-vessel components, to 150 kPa (abs). This is key safety function because a large internal pressure could lead to a breach of the primary confinement barrier. Safety is ensured by discharging the steam evolved during the accident event to the VVPSS suppression tanks where it is condensed. Steam condensation occurs at sub-atmospheric pressure condition. Moreover, being this latter not standard for traditional nuclear systems, this investigation is quite new (not studied in detail before) and deals with an experimental investigation of the direct contact condensation at VVPSS prototypical thermal-hydraulic conditions. To the purpose, a small-scale experimental rig was properly designed and built at Lab. B. Guerrini of DICI-University of Pisa as well as different temperature, pressure and steam mass (flow rate per hole) conditions and sparger patterns have been investigated. The experimental test matrix is also presented in this study. The obtained results show high efficiency of condensation for all examined conditions. The main condensation regimes at sub-atmospheric pressure conditions were identified. In addition, a comparison was done between the condensation regimes experimentally determined and those available in the literature, which were obtained at atmospheric pressure. Finally, results demonstrated to be representative of the real configuration at ITER reactor.

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, ...

The phenomenon of steam-water direct contact condensation (DCC) was widely investigated in the past because of conventional and nuclear engineering applications, like in the safety suppression system of BWRs. When steam is injected into a sub cooled water, different condensation regimes can be observed, such as chugging, condensation oscillation, bubbling condensation oscillation etc. (these have been identified during the experimental campaign on condensation of steam injected into water at sub-atmospheric conditions carried out at Lab. B. Guerrini of DICI - University of Pisa). According to the thermal hydraulic conditions of the condensing jet plume, vibrations may arise. This paper so deals with the analysis of vibration phenomena that originate during the steam-water direct contact condensation at sub-atmospheric conditions, not treated till today. Numerical investigations (by proper FEM code) as well as experimental activity will be presented and discussed in order to evaluate if vibrations are capable to jeopardise this safety suppression system, which is designed to protect Vacuum Vessel of ITER from pressurizing accidents.

Important challenges for fusion technology deal with the design of safety systems aimed to protect the Vacuum Vessel (VV) from pressurizing accidents like the Loss Of Coolant Accident (LOCA). To prevent or mitigate structural damages, the solution proposed is a safety system able to quickly condense released steam in cold water at sub-atmospheric conditions. This water suppression tank (VVPSS) is so aiming at limiting the maximum pressure in the VV to 0.2 MPa during in-vessel coolant leak events and at maintaining the VV long-term pressure below atmospheric pressure during air or incondensable gases ingress, through the Direct Contact Condensation (DCC). The novelty of this study resides especially in the working condition of VVPSS, which operates precisely to sub-atmospheric pressure: up to date no explicit experimental data or investigation of DCC are in fact available in literature. To overcome this lack an extensive experimental work has been done at DICI – University of Pisa, where numerous condensation tests (more than 300) were performed. The operation condition investigated took into account downstream pressure between 30 and 117 kPa and water pool temperature from 30 up to 85 °C. The experimental measurements allow to study the influence of steam mass flux, water temperature and pool pressure on the steam condensation phenomenon (and in turn, based on the stable condensation regime, correctly analyze the design parameter of VVPSS). The results obtained are presented and discussed. Innovative condensation regime maps are in addition provided.

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.

The paper is aimed at preliminary evaluating the load bearing capability of a Generation IV metal cooled reactor (LFR), like the European Lead-cooled System (ELSY), with the intent to support the viability of lead technology for use in a future commercial power plant with waste transmutation capability. Two relevant safety aspects were investigated: the fluid–structure interaction, induced by the propagation of seismic loading (in the event of safe shutdown earthquake), and the high thermal loads due to the residual decay heat (influenced also by the burn up level). The first issue depends on the presence of a liquid free surface that, especially, in SSE condition allows for fluid motion (“sloshing”); the forming and the impact of lead waves could impair the integrity of reactor structures. The second aspect deals with the thermal effects induced by the irradiated core heat source after the reactor shutdown. In order to evaluate the ELSY reactor structural response, induced by both high temperature and hydrodynamic loads, appropriate 3-D finite element models were set up and implemented in MARC code. In addition the input decay power after shutdown has been evaluated by means of MCNPX and Origen2.2 codes. The obtained preliminary results provide information to make possible an upgrading of the reactor design, however with no significant modification of their functionality.

The main goal of the present study is the preliminary evaluation of the seismic demand of a LFR with reference to E uropean L ead Sy stem project (ELSY) considered one of the most promising innovative Generation IV reactor. The safety aspects of the ELSY reactor in the event of a Safe Shutdown Earthquake, taking into account also the effects of the possible fluid–structure interaction, have been analyzed. To the purpose to determine the seismic demand, in according with the international rules, a non-linear dynamic analysis method was used with rather refined 3-D model of LFR for the foreseen structural analyses and simulations of the plant and of the reactor internals behaviour. In this report numerical results are presented and discussed highlighting the relevance of the fluid–structure interaction in terms of structural integrity as well as the isolation technique effectiveness, which is expected to increase the safety margin of the reactor structures during a seismic event, if the isolators frequency is far from that of the reactor. The present work has been performed within the 6th European Framework Project.

Like all industries, the generation of electricity from Nuclear Power Plant produces wastes to be managed. Spent fuel element casks used for transport of nuclear materials must be designed according to rigorous acceptance criteria and standards requirements, e.g. International Atomic Energy Agency ones, in order to provide protection to people and environment against radiation exposure. The aim of this work was the evaluation of the integrity of spent fuel cask under both normal and accident transport conditions, such as impact (9 m drop impact event onto a flat, essentially unyielding, horizontal surface, in the most damaging orientation) and rigorous fire events (full exposure to an engulfing fire for 30 min (fire test) or to an environment at 800 °C temperature for a numerical simulation or for a furnace test). Using the finite element code ANSYS both steady-state and transient thermal analyses were carried out to determine the maximum fuel temperature and the temperatures behaviour into the cask, considering all the heat transfer modes between the cask and the external environment as well as inside the cask itself. Moreover, both wet and dry fuel storage inside the cavity of the body were analyzed. The obtained results, used for the new licensing approval by the Italian competent Authority of the cask for PWR spent fuel transport, are discussed.

Integrity of a Nuclear Power Plant (NPP) must be ensured during the plant life in any design condition and, particularly, in the event of a severe earthquake. To investigate the seismic resistance capability of as-built structures systems and components, in the event of a Safe Shutdown Earthquake (SSE), and analyse its related effects on a near term deployment reactor and its internals, a deterministic methodological approach, based on the evaluation of the propagation of seismic waves along the structure, was applied considering, also, the use of innovative anti-seismic techniques. In this paper the attention is focused on the use and influence of seismic isolation technologies (e.g. isolators based on passive energy dissipation) that seem able to ensure the full integrity and operability of NPP structures, to enhance the seismic safety (improving the design of new NPPs and if possible, to retrofit existing facilities) and to attain a standardization plant design. To the purpose of this study a numerical assessment of dynamic 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) are herein presented and compared in order to highlight the isolation technique effectiveness.

The seismic response analysis of such liquid storage systems, especially liquid metal reactors, as for example the eXperimental Accelerator Driven System (XADS), was examined taking into account mainly the coupling effects of the fluid–structure interaction and their influence on its relevant internal systems and components. Therefore this paper deals with the structural analyses of the seismically induced hydrodynamic responses, in the event of a safe shutdown earthquake (SSE), and the free oscillation (known as sloshing waves) of a metal liquid coolant as well as the dynamic buckling effects on involved structures. To the mentioned purpose the interaction and coupling effects among the main reactor vessel structures and the primary coolant response were investigated by means of a numerical evaluation (with a qualified finite element code) because of the lack of analytical linear theories that in any case are not adequate to describe all the complex phenomena related to the seismic loading. For the numerical modelling procedure, 3D finite element models were set up to analyse the propagation of seismic waves as well as its derived structural effects, such as the fluid steep waves motion, the local buckling bulges, etc., taking into account the geometrical and material nonlinearities of the RPV and the considered simplified internals. The obtained numerical results in terms of stress intensity and of the capability of the structures to resist relevant seismic loads are, thus, presented and discussed. Moreover the performed analyses allowed to highlight the structures mostly affected by the assumed loading conditions in order to achieve data useful for an upgrading of the design geometry, if any, for the considered reactor.

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.

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.

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.

This paper deals with the treatment of methods for a preliminary safety assessment of next generation nuclear power plant (of Liquid Metal cooled Fast Reactor types) structures. As an example, it was considered the European Lead-cooled System (ELSY) international project response under a reference design basis earthquake (Safe Shutdown Earthquake-SSE) excitation, taking into account also the main isolating device effects in order to increase the reactor safety. This study is intended to analyze the structural effects of a reference SSE by means of an appropriate dynamic FEM code, for possible geometries of both not isolated and isolated ELSY containment building foundation cases. A great attention should be focused too on the arisen hydrodynamic forces and coupling effects between fluid and structures (sloshing phenomenon) that may impair the structures resistance and/or operating capabilities. The obtained numerical results were critically analysed with the intent also to contribute to a step of the safety optimization of the mentioned systems.

The casks used for transport of nuclear materials, especially the spent fuel element (SPE), must be designed according to rigorous acceptance criteria and standards requirements, e.g. the International Atomic Energy Agency ones, in order to provide protection to people and environment against radiation exposure particularly in a severe accident scenario. The aim of this work was the evaluation of the integrity of a spent fuel cask under both normal and accident scenarios transport conditions, such as impact and rigorous fire events, in according to the IAEA accident test requirements. The thermal behaviour and the temperatures distribution of a Light Water Reactor (LWR) spent fuel transport cask are presented in this paper, especially with reference to the Italian cask designed by AGN, which was characterized by a cylindrical body, with water or air inside the internal cavity, and two lateral shock absorbers. Using the finite element code ANSYS a series of thermal analyses (steady-state and transient thermal analyses) were carried out in order to obtain the maximum fuel temperature and the temperatures field in the body of the cask, both in normal and in accidents scenario, considering all the heat transfer modes between the cask and the external environment (fire in the test or air in the normal conditions) as well as inside the cask itself. In order to follow the standards requirements, the thermal analyses in accidents scenarios were also performed adopting a deformed shape of the shock absorbers to simulate the mechanical effects of a previous IAEA 9 m drop test event. Impact tests on scale models of the shock absorbers have already been conducted in the past at the Department of Mechanical, Nuclear and Production Engineering, University of Pisa, in the ‘80s. The obtained results, used for possible new licensing approval purposes by the Italian competent Authority of the cask for PWR spent fuel cask transport by the Italian competent Authority, are discussed.

Nuclear power plant (NPP) design is strictly dependent on seismic hazard and safety aspects concerned with the external events of the site. Earthquake resistant structures design requires realistic and accurate physical and theoretical models to describe the response of the nuclear power plants (NPPs) that depend on both the ground motion characteristics and the dynamic properties of the structures themselves. In order to improve the design of new NPPs and, at the same time, to retrofit existing ones the dynamic behaviour of structures subjected to critical seismic excitations that may occur during their expected service life must be evaluated. The aim of this work is to select new effective methods to assess NPPs vulnerability by properly capturing the effects of a safe shutdown earthquake (SSE) event on nuclear structures, like the near term deployment IRIS reactor, and to evaluate the seismic resistance capability of as-built structures systems and components. To attain the purpose a validated deterministic methodology based on an accurate finite element modelling coupled to substructure and time history approaches was employed for studying the overall dynamic behaviour of the NPP relevant components. Moreover the set up three-dimensional model was also validated to evaluate the performance and reliability of the adopted FEM code (mesh refinements and type element influence). This detailed numerical assessment, involving the most widely used finite element numerical codes (MSC.Marc ® and Ansys ® ), allowed to solve, perform and simulate as accurately as possible the dynamic behaviour of structures which may withstand a lot of more or less complicate structural problems. To evaluate the accuracy and the reliability as well as to determine the related error of the set-up procedure, the obtained seismic analyses results in term of accelerations, propagated from the ground to the auxiliary building systems and components, and displacements were compared highlighting a very good agreement.

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 © 2009 by ASME

The present paper deals with instabilities of long homogeneous and isotropic thin shells, characterized by geometrical non-linearities and imperfections, with reference to a bent helicoidally geometrical shape of particular interest for the helicoidal steam generators tube bundle. Apparently no data exist in the literature to describe the non-linear buckling behaviour of curved thin shells under external pressure, thus, the theoretical analyses based on the classical linear elastic theory, as expected, might be inadequate to evaluate the collapse load especially if the curvature is rather large. To the purpose of determining the buckling pressure load the effects of a pre-existing level of geometrical and technological imperfection, unavoidably caused by various manufacturing processes were also considered. A numerical analysis, based on detailed three-dimensional finite element (FE) code, was performed on curved thin shell subject to combined pressure and in-plane bending in order to take into account the non-linear geometrical and material properties as well as to assess and quantify their detrimental effects on the buckling phenomenon behaviour. Moreover, at Pisa University a rather intense experimental research activity is being carried out on the buckling of thin-walled tube specimens in the dimensional range suitable for the above mentioned applications. Collapse experiments have demonstrated that the detriment of the mechanical properties resulted in a reduction in the collapse pressure load of about 15% and the obtained results have highlighted also the dependence of the buckling behaviour on the stress-initial defects/imperfection width (mainly on the diameter-to-thickness or tube/bending diameters ratios) relationship.

The load carrying behaviour of cylindrical thin-walled shell structures under pressure load is strongly dependent on the nature and magnitude of the imperfections invariably caused by various manufacturing processes. The present paper examines instabilities of long homogeneous and isotropic thin elastic tubes, characterized by geometric imperfections like eccentricity or ovality, on the buckling behaviour in conditions for which, at present, a complete theoretical analysis was not found in literature. Moreover, the additional aspect of the influence of the welded joint geometry and position is investigated over a wide range of diameter to thickness ratio, extending the findings of previous works. The problem of buckling for variable load conditions is relevant in the context of NPP applications as, for instance the optimisation of an integrated and innovative LWR Steam Generator (SG) tubes, according to the updated ASME rules. To the purpose, at Pisa University a rather intense research activity is being carried out on the buckling of thin walled metal specimens in the dimensional range suitable for the above mentioned application. Therefore a test equipment (with the necessary data acquisition facility), suitable for carrying out test series on this issue, as well as numerical models implemented on the MARC FEM code, were set up. The experiments were conducted on test specimens with different materials, e.g. A-316 ASTM (with and without seam weld) and Inconel 690 TT, as well as different loading conditions (lateral and hydrostatic external pressure). A validation of numerical evaluations by comparison with test results is also performed. A good agreement has been observed between the experimental data and the elasto-plastic finite element analyses results, highlighting also the different influence of the mentioned imperfections on the buckling loads.

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 © 2008 by ASME

The evaluation of the thermal conductivity of breeder materials is one of the main goals to find the best candidate material for the fusion reactor technology.The aim of this paper is to evaluate experimentally the thermal conductivity of a ceramic material by applying the hot wire method at different temperatures, ranging from 50 to about 800°C.The updated experimental facility, available at the Department of Civil and Industrial Engineering (DICI) of the University of Pisa, used to determine the thermal conductivity of a ceramic material (alumina), will be described along with the measurement acquisition system. Moreover it will be also provided an overview of the current state of art of the ceramic pebble bed breeder thermos-mechanics R&D (e.g. Lithium Orthosilicate (Li4SiO4) and Lithium Metatitanate (Li2TiO3)) focusing on the up-to-date analysis.The methodological approach adopted is articulated in two phase: the first one aimed at the experimental evaluation of thermal conductivity of a ceramic material by means of hot wire method, to be subsequently used in the second phase that is based on the test rig method, through which is measured the thermal conductivity of pebble bed material. In this framework, the experimental procedure and the measured results obtained varying the temperature, are presented and discussed.

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.

This paper illustrates the preliminary experimental tests for determining the effective thermal conductivity of ceramic pebble beds versus temperature and compression strains. Ceramic pebble beds are promising candidates to be used in breeding blankets for nuclear fusion reactor as breeder and neutron multiplier. The tests were performed with an experimental rig, built at the DICI-University of Pisa, which permits to determine the thermal conductivity of pebble beds in steady state conditions, at several temperatures and compression forces. The values of thermal conductivity of pebble beds are obtained as function of a known conductivity of an alumina disc. The assessment of the method has been performed determining the effective thermal conductivity of alumina pebbles beds of different diameters. Void fraction and compression strains are the parameters that mainly influence the variability of the thermal conductivity of the beds.