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

Analysis of feasibility of a new core catcher for the in-vessel core melt retention strategy

This study deals with the feasibility study of a new in-vessel core melt retention (IVCMR) strategy capable to extend the coping period in the event of adverse situations, involving the melting of the core. Since Fukushima accident, many studies have been carried out to resolve the severe accident mitigation issues related to the corium stabilization inside and outside the reactor vessel. This is in fact one of the most relevant safety issues to secure LWRs from the point of view of severe accident mitigation and containment integrity. As for the corium stabilization inside the reactor vessel, in this study it is proposed a new IVCMR concept, developed at the University of Pisa, based on the adoption of an original core catcher design made of batches of ceramic material. By profiting of its low thermal conductivity, this core catcher is capable to retard the heat-up of the lower head of the vessel during the phase of relocation of the corium. To support the feasibility of its design analytical and numerical analyses have been performed assuming homogeneous pool condition. Results show that the adoption of the proposed core catcher solution extends the severe accident coping period: after 1 h from the initiating event, the maximum temperature of the vessel wall is below the limit for which localized failure may appear

ITER cryostat accidental scenario: fluid dynamics analysis of Ingress of Coolant Event Accident

ITER (International Thermonuclear Experimental Reactor) is an experimental reactor aimed at demonstrating the technological and scientific feasibility of fusion technology. A future fusion power plant producing large amounts of energy power will be required to breed all of its own Tritium. ITER will demonstrate this essential concept of Tritium self-sustainment. Among the most important components of that reactor there is the cryostat that is, specifically, a large stainless steel structure surrounding the vacuum vessel and the superconducting magnets, providing a super-cool vacuum environment. The aim of this paper is to evaluate the effects caused by a suddenly rupture of one of the cryogenic lines with release of helium inside the cryostat, event known as CrICE: Ingress of Coolant Event in Cryostat. The CrICE accident scenario has been simulated by ANSYS©CFX. To the purpose, a suitable model representing a 20° sector of the overall ITER structures, vacuum vessel, magnets, thermal shield, ports and cryostat was set up and implemented, in order to characterize and define the free volume to be filled by the gas that would be released eventually as well as the air inside the bioshield. The numerical model, the geometrical characteristics and the materials properties used as input in the simulation of the accidental scenario have been presented and discussed. The results obtained indicated that the cryostat is capable to sustain the pressure and the thermal loads generated by the accident conditions. It is also worthy to remark that these results (raw outcomes) will be used for a further detailed investigation of the structural performances of cryostat itself

EM zooming procedure in ANSYS Maxwell 3D

The severity of electromagnetic (EM) loads produced by plasma disruptions is one of the most concerning issues for the ITER in-vessel components design. To investigate the effects of fast EM transients on plasma surrounding structures during a disruption the Secondary Excitations (SE) method is used. This is an interface procedure to couple 2D plasma equilibrium codes with Finite Elements (FE) software. The Zooming Approach (ZA) used for the analyses presented here is a particular implementation of the SE method. The aim of this work is the demonstration that the ZA can be effectively applied in case of ANSYS Maxwell 3D analyses combining the ease of use of the Maxwell code with the computational efficiency of the ZA. The work has been carried out evaluating the EM loads acting on the ITER Diagnostic Equatorial Port Plug (EPP) during major disruptions scenario and comparing these loads with those obtained in previous analyses. Additional analyses have been performed to study the effect of ferromagnetic materials on EM loads in order to investigate ANSYS Maxwell capabilities in simulating non-linear magnetic properties

Numerical evaluation of sloshing effects in ELSY innovative nuclear reactor pressure vessels seismic response

Paper presented at the 6th International Conference on Heat Transfer, Fluid Mechanics and Thermodynamics, South Africa, 30 June - 2 July, 2008.In Europe a great effort has been made in the Lead-Bismuth Eutectic (LBE) technology, for use in the sub-critical reactors, and its natural development is represented by the use of pure lead that is less corrosive, chemically inert and in the foreseen environment has good neutronic and thermal-hydraulic characteristics, therefore it appears to be a suitable coolant for a fast reactor. The main purpose of this study deals with the evaluation of the sloshing dynamic effects of lead coolant during a safety shut down earthquake applied to a conceptual Lead-cooled Fast Reactor (LFR) Generation IV (GEN IV) Nuclear Power Plant design, with reference to the ELSY project system configuration that is under development within the ongoing European 7FW ELSY Program. ELSY is an innovative small size pool-type reactor (600 MWe) cooled by pure lead, characterized by a compact and simple integrated primary circuit; by the way this configuration is favourable from the point of view of the reduction of the seismic loads and of the negative effect of the high lead density. Therefore, the fluid-structure interaction problems and the free oscillations of the heavy metal primary coolant attracted the attention because during a strong motion earthquake the lead surrounding the internals may be accelerated and the so-called hydrodynamic interaction, due to the coolant sloshing, may significantly influence the stress level in the reactor pressure vessel (RPV). To the purpose, the effect of the rigidity of adjacent internals walls and coupling between coolant and vessel are considered An adequate numerical modelling, by means a 3-D finite element model, was set up and used for the foreseen structures dynamic analysis, due to the inability of linear theory to describe accurately the wave’s motion accounting for the complex considered RPV geometrical aspect as well as the material nonlinearities. Numerical results are presented and discussed highlighting the importance of the fluid-structure interaction effects in terms of stress intensity as well as the capacity of internals and vessel walls to withstand wave’s impact and prevent instabilities.vk201

Unsupervised anomaly detection in pressurized water reactor digital twins using autoencoder neural networks

Deep learning (DL), that is becoming quite popular for prediction and analysis of complex patterns in large amounts of data is used to investigate the safety behaviour of the nuclear plant items. This is achieved by using multiple layers of artificial neural networks to process and transform input data, allowing for the creation of highly accurate predictive models. Particularly to the aim the unsupervised machine learning approach and the digital twin concept in form of pressurized water reactor 2-loop simulator are used. This innovative methodology is based on neural network algorithm that makes capable to predict failures of plant structure, system, and components earlier than the activation of safety and emergency systems. Moreover, to match the objective of the study several scenarios of loss of cooling accident (LOCA) of different break size were simulated. To make the acquisition platform realistic, Gaussian noise was added to the input signals. The neural network has been fed by synthetic dataset provide by PCTRAN simulator and the efficiency in event identification was studied. Further, due to the very limited studies on the unsupervised anomaly detection by means of autoencoder neural networks applied for plant monitoring and surveillance, the methodology has been validated with experimental data from resonant test rig designed for fatigue testing of tubular components. The obtained results demonstrate the reliability and the efficiency of the methodology in detecting anomalous events prior the activation of safety system. Particularly, if the difference between the expected readings and the collected data goes beyond the predetermined threshold, then the anomalous event is identified, e.g., the model detected anomalies up to 38 min before the reactor scram intervention

Preliminary analysis of an aged RPV subjected to station blackout

Today, 46% of operating Nuclear Power Plants (NPP) have a lifetime between 31 and 40 years, while 19% have been in operation for more than 40 years. Long Term Operation (LTO) is an urgent requirement for all of the nuclear industry. The aim of this study is to assess the performance of a reactor pressure vessel (RPV) subjected to a station blackout (SBO) event. Alterations suffered by the material properties and creep at elevated temperatures are considered. In this study, coupling between MELCOR and Finite Element Method (FEM) codes is carried out. In the Finite Element (FE) model, the combined effects of ageing and creep are implemented through degraded material properties and a viscoplastic model. The reliability of the model is validated by comparing the FOREVER/C1 experimental results. The results show that the RPV lower head bends downwards with a maximum radial expansion of about 260 mm and RPV thermomechanical properties are reduced by more than 50% at high temperatures. The effects of ageing, creep and long heat-up strongly affect the resistance of the RPV system until the point of compromising it in the absence of/delayed emergency intervention. Aged RPV at end-of-life may collapse earlier, and in less time, with the same accidental conditions

Performance of a type IP-2 packaging system in accident conditions of transport

Paper presented to the 10th International Conference on Heat Transfer, Fluid Mechanics and Thermodynamics, Florida, 14-16 July 2014.A package to be used for the transport of hazardous /radioactive materials must demonstrate to fulfil the International standards requirements in order to provide protection to the human being and environment even under accident conditions, such as rigorous fire events. In these conditions, the system (package or cask), constituted, in general, by a massive sealed steel vessel, must thus demonstrate to be robust, safe and reliable so to guarantee both structural strength and radiation shielding. The present study deals with the evaluation of the thermostructural response and performance of an Italian design type IP-2 packaging system, provided by Sogin, that should be adopted for the transportation of low and intermediate level radioactive solid/solidified wastes. To evaluate its performance, a FEM model has been set up and implemented in a rather refined way taking into account all the packaging system components. Numerical simulations addressed fire scenarios as specified in the IAEA regulations: packaging subjected to an engulfing fire of 800 °C for 30 minutes. All the heat transfer mechanisms, inside the system and between the system itself and the environment, have been considered in the thermal analyses performed. The results of the thermal analyses are presented and discussed. Analysing the results obtained it is possible to conclude that although any potential damage the integrity of IP2 packaging system is assured.cf201

experimental investigation of functional performance of a vacuum vessel pressure suppression system of iter

Abstract 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

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