benefits of seismic isolation for nuclear structures subjected to severe earthquake

The Fukushima accident has reiterated that the seismic safety is a clear necessity in the design of nuclear power plants. To overcome the weaknesses of the plant design, appropriate measures or interventions have thus to be put in place to improve the nuclear safety. In this study, seismic isolation, widely adopted for conventional constructions, is considered as retrofit measure to provide superior performance of plant itself, even when exceptional events occur. In this paper, we numerically investigate the dynamic behaviour of a Small Modular Reactor (SMR) plant subjected to 0.6g PGA; in doing that time-history analysis has been performed assuming the reactor building with and without isolators. For that purpose, a suitable FEM model has been implemented to provide in-structure response spectra at safety relevant locations and subsystem supports. Adequate steel and concrete properties as well as isolators properties, experimentally determined, have been assumed. Results have shown the benefits of seismic isolation for NPP that can so sustain levels of loading beyond the design input and demonstrated that failure of an isolation system cannot occur before failure of the isolated structure. However, the large horizontal displacements of the structure require appropriate considerations in the layout and interfaces for interconnected systems

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

Preliminary investigation of Li4SiO4 pebbles structural performance

Abstract One of the main purposes of a breeder blanket is to remove the heat produced in the blanket by the fusion reaction neutrons, and to breed the tritium required to sustain it. To achieve these requirements, several breeder materials (solid or liquid lithium-bearing ones) have been investigated in the past decades. To date it has not yet been possible to identify a stable material, with high thermal conductivity and melting point. This paper deals with the mechanical characterization of the lithium orthosilicate (Li4SiO4) in form of pebbles, produced at the University of Pisa at room temperature by a drip casting forming technique, starting from an aqueous suspension of Li4SiO4 precursor prepared by a sol-gel synthesis method. To investigate also numerically, by means of FE code, the breeder blanket behaviour, it is of meaningful importance the mechanical characterization of such pebbles. To the purpose, either static or cyclic uniaxial compression tests, without radial constraints, have been performed on several produced pebbles of about 1.5 mm diameter in order to determine the collapse and crushing loads and the stiffness. Moreover, the carried-out post-test SEM examination allowed to evaluate the failure mode and the crack shapes on the contact surface. Results show the influence of the elastic properties and matrix flaw population on the crushing load. The pebbles produced by the sol-gel method showed also a high strength, the value of which is comparable to that of the pebbles obtained by melting process

Demonstration of structural performance of IP-2 package by simulation and full-scale horizontal free drop test

Packaging systems for the transportation of radioactive wastes have to be designed according to rigorous acceptance criteria and requirements in order to protect people and environment against radiation exposure and contamination risk. The IAEA requirements for type IP2 (Industrial Package Type 2) packages include to carry out free drop tests that represent normal conditions of transport. In such conditions, obviously, the required containment capability of the package has to be ensured. In this study the mechanical performances of a new Italian packaging system for the transportation of low and intermediate level wastes (LILW) undergoing horizontal free drop test are investigated. Especially, deformations caused in the sealing area of the package, which can affect the capability of the containment system, are evaluated. The carried out numerical analyses and experimental tests, at the lab. Scalbatraio of the DICI- University of Pisa, are presented and discussed. Numerical analyses (by qualified MARC® code) have been performed to investigate the stress histories in the bolts, lid, and package body as well as the deformations in the sealing area and the compression conditions of the gasket. Localised stress appeared at the flange and at the bottom of packaging system. The maximum stresses resulted lower than the stress limits, so the structural integrity of the package was maintained and confirming its tightness. As a consequence of the primary impact a local deformation appeared at the primary lid, no cliff edge or loss of the safety features resulted

Structural performance of an IP2 package in free drop test conditions: numerical and experimental evaluations

The casks or packaging systems used for the transportation of nuclear materials, especially spent fuel elements, have to be designed according to rigorous acceptance requirements, like the IAEA ones, in order to provide protection to human beings and environment against radiation exposure and contamination. This study deals with the free drop test of an Italian design packaging system to be used for the transportation of low and intermediate level radioactive wastes. Impact drop experiments were performed in the Lab. Scalbatraio of the DICI - University of Pisa. Dynamic analyses too have been carried out, by refined models of both the cask and target surface to predict the effects of the impact shock (vertical drop) on the package. The experimental tests and numerical analyses are thoroughly compared, presented and discussed. The numerical approach shows to be suitable to reproduce with good reliability the test situations and results

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

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

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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

Characterization of the thermal conductivity for ceramic pebble beds

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