Simultaneous Nasopharyngeal Carriage of Two Pneumococcal Multilocus Sequence Types with a Serotype 3 Phenotype

Knowledge of the epidemiology of pneumococcal disease in Bolivia is sparse, and Multilocus Sequence Typing (MLST) of isolates has not been previously possible. Beni state has until recently been a geographically isolated region of the Bolivian Amazon basin and is a region of significant poverty. During June and July 2007, we performed a pneumococcal carriage study recruiting over 600 schoolchildren in two towns in the Beni state. Here, we describe the unique identification of simultaneous nasopharyngeal carriage of two pneumococcal multilocus sequence types with a serotype 3 phenotype within a single subject

Analytische Modellierung mechanischer Schwingungen von Primärkreiskomponenten des Druckwasserreaktors WWER-440 mit finiten Elementen

The project contributes to the improved evaluation of the mechanical integrity of the soviet-type VVER-440 reactors especially, to a sensitive early failure detection and to the localization of mechanical damages of reactor components by means of vibration monitoring. For that purpose the mechanical vibration of all primary circuit components was modelled by finite elements. Modeling was built on the finite element code ANSYS. The interaction between the coolant flowing in the downcomer and the vibrating components has been considered by a fluid-structure element, which describes additional mode selective damping and intertia due to the coolant displacement when the downcomer geometry changes. The calculation model was adjusted using results from experimental vibration investigations. To some extent data from earlier measurements were available. But additionally dedicated experiments had to be performed at original VVERs. Now, the model can be regarded to be widely verified. Mainly it was applied to clarify how hypothetical damages of reactor internals influence the vibration signature of the primary circuit. Such kind of damage simulation is an appropriate means to find sensitive measuring positiones for on-line monitoring and to define physically based threshold values. In principle, the model is even suited to estimate the loads of reactor components which might be imposed by external events (explosion, earthquake)

Modelling of in-vessel retention after relocation of corium into the lower plenum

Considering the unlikely core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be investigated for a determination of the loadings on the containment. Worldwide several experiments have been performed accompanied with material properties evaluation, theoretical, and numerical work. At the Institute of Safety Research of the FZR a finite element model has been de-veloped simulating the thermal processes and the viscoplastic behaviour of the ves-sel wall. An advanced model for creep and material damage has been established and has been validated using experimental data. The thermal and the mechanical calculations are sequentially and recursively coupled. The model is capable of evalu-ating fracture time and fracture position of a vessel with an internally heated melt pool. The model was applied to pre- and post test calculations for the FOREVER test se-ries representing the lower head RPV of a PWR in the geometrical scale of 1:10. These experiments were performed at the Royal Institute of Technology in Stock-holm. The results of the calculations can be summarised as follows: # The creeping process is caused by the simultaneous presence of high tem-perature (>600 °C) and pressure (>1 MPa) # The hot focus region is the most endangered zone exhibiting the highest creep strain rates. # The exact level of temperature and pressure has an influence on the vessel failure time but not on the failure position # The failure time can be predicted with an uncertainty of 20 to 25%. This uncer-tainty is caused by the large scatter and the high temperature sensitivity of the viscoplastic properties of the RPV steel. # Contrary to the hot focus region, the lower centre of the vessel head exhibits a higher strength because of the lower temperatures in this zone. The lower part moves down without significant deformation. Therefore it can be assumed, that the vessel failure can be retarded or prevented by supporting this range. # The development of a gap between melt crust and vessel wall could not be proofed. First calculations for a PWR geometry were performed to work out differences and commonalities between prototypic scenarios and scaled experiments. The results of the FOREVER-experiments cannot be transferred directly to PWR geometry. The geometrical, mechanical and thermal relations cannot be scaled in the same way. Because of the significantly higher temperature level, a partial ablation of the vessel wall has to be to expected in the PWR scenario, which is not the case in the FOREVER tests. But nevertheless the FOREVER tests are the only integral in-vessel retention experiments up to now and they led to a number of important insights about the behaviour of a vessel under the loading of a melt pool and pressure

Energiedispersive Untersuchung der Wechselwirkung schneller Neutronen mit Materie; Teilbericht : Auslegung des Neutronen-Produktionstargets

Der Aufbau und die erste Nutzung eines kompakten Flugzeitsystems zur energiedispersiven Untersuchung der Wechselwirkung von schnellen Neutronen mit Materialien sind Inhalt eines Vernetzungsprojektes des Forschungszentrums Rossendorf, an dem auch die Technische Universität Dresden im Rahmen eines gemeinsamen DFG-Projektes mitarbeitet. Die geplanten Flugzeit-Experimente mit gepulsten Neutronen werden an der Strahlungsquelle ELBE durchgeführt werden. Erste Ergebnisse zur Entwicklung eines Neutronen-Produktionstargets werden vorgestellt. Mit Hilfe von Strahlungstransport- und Finite-Elemente-Programmen wurden die Verteilungen der Energiefreisetzung des von der Strahlungsquelle ELBE genutzten Elektronenstrahls und der Temperatur im Neutronen-Radiator sowie die zu erwartenden Teilchenspektren und -flüsse am Messplatz berechnet. Überlegungen zur Entwicklung des Strahlfängers werden diskutiert

Analytische Modellierung mechanischer Schwingungen von Primärkreiskomponenten des Druckwasserreaktors WWER-440 mit finiten Elementen

The project contributes to the improved evaluation of the mechanical integrity of the soviet-type VVER-440 reactors especially, to a sensitive early failure detection and to the localization of mechanical damages of reactor components by means of vibration monitoring. For that purpose the mechanical vibration of all primary circuit components was modelled by finite elements. Modeling was built on the finite element code ANSYS. The interaction between the coolant flowing in the downcomer and the vibrating components has been considered by a fluid-structure element, which describes additional mode selective damping and intertia due to the coolant displacement when the downcomer geometry changes. The calculation model was adjusted using results from experimental vibration investigations. To some extent data from earlier measurements were available. But additionally dedicated experiments had to be performed at original VVERs. Now, the model can be regarded to be widely verified. Mainly it was applied to clarify how hypothetical damages of reactor internals influence the vibration signature of the primary circuit. Such kind of damage simulation is an appropriate means to find sensitive measuring positiones for on-line monitoring and to define physically based threshold values. In principle, the model is even suited to estimate the loads of reactor components which might be imposed by external events (explosion, earthquake)

Modelling of in-vessel retention after relocation of corium into the lower plenum

Considering the unlikely core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be investigated for a determination of the loadings on the containment. Worldwide several experiments have been performed accompanied with material properties evaluation, theoretical, and numerical work. At the Institute of Safety Research of the FZR a finite element model has been de-veloped simulating the thermal processes and the viscoplastic behaviour of the ves-sel wall. An advanced model for creep and material damage has been established and has been validated using experimental data. The thermal and the mechanical calculations are sequentially and recursively coupled. The model is capable of evalu-ating fracture time and fracture position of a vessel with an internally heated melt pool. The model was applied to pre- and post test calculations for the FOREVER test se-ries representing the lower head RPV of a PWR in the geometrical scale of 1:10. These experiments were performed at the Royal Institute of Technology in Stock-holm. The results of the calculations can be summarised as follows: # The creeping process is caused by the simultaneous presence of high tem-perature (>600 °C) and pressure (>1 MPa) # The hot focus region is the most endangered zone exhibiting the highest creep strain rates. # The exact level of temperature and pressure has an influence on the vessel failure time but not on the failure position # The failure time can be predicted with an uncertainty of 20 to 25%. This uncer-tainty is caused by the large scatter and the high temperature sensitivity of the viscoplastic properties of the RPV steel. # Contrary to the hot focus region, the lower centre of the vessel head exhibits a higher strength because of the lower temperatures in this zone. The lower part moves down without significant deformation. Therefore it can be assumed, that the vessel failure can be retarded or prevented by supporting this range. # The development of a gap between melt crust and vessel wall could not be proofed. First calculations for a PWR geometry were performed to work out differences and commonalities between prototypic scenarios and scaled experiments. The results of the FOREVER-experiments cannot be transferred directly to PWR geometry. The geometrical, mechanical and thermal relations cannot be scaled in the same way. Because of the significantly higher temperature level, a partial ablation of the vessel wall has to be to expected in the PWR scenario, which is not the case in the FOREVER tests. But nevertheless the FOREVER tests are the only integral in-vessel retention experiments up to now and they led to a number of important insights about the behaviour of a vessel under the loading of a melt pool and pressure

Analytische Modellierung mechanischer Schwingungen von Primärkreiskomponenten des Druckwasserreaktors WWER-440 mit finiten Elementen

The project contributes to the improved evaluation of the mechanical integrity of the soviet-type VVER-440 reactors especially, to a sensitive early failure detection and to the localization of mechanical damages of reactor components by means of vibration monitoring. For that purpose the mechanical vibration of all primary circuit components was modelled by finite elements. Modeling was built on the finite element code ANSYS. The interaction between the coolant flowing in the downcomer and the vibrating components has been considered by a fluid-structure element, which describes additional mode selective damping and intertia due to the coolant displacement when the downcomer geometry changes. The calculation model was adjusted using results from experimental vibration investigations. To some extent data from earlier measurements were available. But additionally dedicated experiments had to be performed at original VVERs. Now, the model can be regarded to be widely verified. Mainly it was applied to clarify how hypothetical damages of reactor internals influence the vibration signature of the primary circuit. Such kind of damage simulation is an appropriate means to find sensitive measuring positiones for on-line monitoring and to define physically based threshold values. In principle, the model is even suited to estimate the loads of reactor components which might be imposed by external events (explosion, earthquake)

Modelling of in-vessel retention after relocation of corium into the lower plenum

Considering the unlikely core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be investigated for a determination of the loadings on the containment. Worldwide several experiments have been performed accompanied with material properties evaluation, theoretical, and numerical work. At the Institute of Safety Research of the FZR a finite element model has been de-veloped simulating the thermal processes and the viscoplastic behaviour of the ves-sel wall. An advanced model for creep and material damage has been established and has been validated using experimental data. The thermal and the mechanical calculations are sequentially and recursively coupled. The model is capable of evalu-ating fracture time and fracture position of a vessel with an internally heated melt pool. The model was applied to pre- and post test calculations for the FOREVER test se-ries representing the lower head RPV of a PWR in the geometrical scale of 1:10. These experiments were performed at the Royal Institute of Technology in Stock-holm. The results of the calculations can be summarised as follows: # The creeping process is caused by the simultaneous presence of high tem-perature (>600 °C) and pressure (>1 MPa) # The hot focus region is the most endangered zone exhibiting the highest creep strain rates. # The exact level of temperature and pressure has an influence on the vessel failure time but not on the failure position # The failure time can be predicted with an uncertainty of 20 to 25%. This uncer-tainty is caused by the large scatter and the high temperature sensitivity of the viscoplastic properties of the RPV steel. # Contrary to the hot focus region, the lower centre of the vessel head exhibits a higher strength because of the lower temperatures in this zone. The lower part moves down without significant deformation. Therefore it can be assumed, that the vessel failure can be retarded or prevented by supporting this range. # The development of a gap between melt crust and vessel wall could not be proofed. First calculations for a PWR geometry were performed to work out differences and commonalities between prototypic scenarios and scaled experiments. The results of the FOREVER-experiments cannot be transferred directly to PWR geometry. The geometrical, mechanical and thermal relations cannot be scaled in the same way. Because of the significantly higher temperature level, a partial ablation of the vessel wall has to be to expected in the PWR scenario, which is not the case in the FOREVER tests. But nevertheless the FOREVER tests are the only integral in-vessel retention experiments up to now and they led to a number of important insights about the behaviour of a vessel under the loading of a melt pool and pressure

Simultaneous Nasopharyngeal Carriage of Two Pneumococcal Multilocus Sequence Types with a Serotype 3 Phenotype

Knowledge of the epidemiology of pneumococcal disease in Bolivia is sparse, and Multilocus Sequence Typing (MLST) of isolates has not been previously possible. Beni state has until recently been a geographically isolated region of the Bolivian Amazon basin and is a region of significant poverty. During June and July 2007, we performed a pneumococcal carriage study recruiting over 600 schoolchildren in two towns in the Beni state. Here, we describe the unique identification of simultaneous nasopharyngeal carriage of two pneumococcal multilocus sequence types with a serotype 3 phenotype within a single subject

Energiedispersive Untersuchung der Wechselwirkung schneller Neutronen mit Materie; Teilbericht : Auslegung des Neutronen-Produktionstargets

Der Aufbau und die erste Nutzung eines kompakten Flugzeitsystems zur energiedispersiven Untersuchung der Wechselwirkung von schnellen Neutronen mit Materialien sind Inhalt eines Vernetzungsprojektes des Forschungszentrums Rossendorf, an dem auch die Technische Universität Dresden im Rahmen eines gemeinsamen DFG-Projektes mitarbeitet. Die geplanten Flugzeit-Experimente mit gepulsten Neutronen werden an der Strahlungsquelle ELBE durchgeführt werden. Erste Ergebnisse zur Entwicklung eines Neutronen-Produktionstargets werden vorgestellt. Mit Hilfe von Strahlungstransport- und Finite-Elemente-Programmen wurden die Verteilungen der Energiefreisetzung des von der Strahlungsquelle ELBE genutzten Elektronenstrahls und der Temperatur im Neutronen-Radiator sowie die zu erwartenden Teilchenspektren und -flüsse am Messplatz berechnet. Überlegungen zur Entwicklung des Strahlfängers werden diskutiert