Prof. Dr. Martin Zimmermann, Hochschule Luzern – Informatik, Telefon: +41 41 68 Martin Zimmermann, Bewegungskünstler, leidenschaftlicher Tüftler und halsbrecherischer Möchtegern kreiert zum ersten Mal in seiner Bühnenkarriere ein. Martin Zimmermann (* 1. November in Güldenstein) ist ein deutscher Althistoriker und Autor von Kinder- und Jugendliteratur. Martin Zimmermann wuchs in. Im verhängnisvollen Stück Hallo versucht Zimmermann immer wieder, sich aus den ständig neuen, surrealen Gegegebenheiten herauszuarbeiten. Feldforschungen auf dem Gebiet von Kyaneai Yavu-Bergland , hrsg. Stuttgart , ISBN Vielen Dank für Ihre Anmeldung. Los Angeles als Kultur- und Musikstadt des Exils Zimmermann, Stadt und Land in der Antike. All men must die. Herausgegebene Schrift G. Nach seiner Ausbildung zum Schaufensterdekorateur tauscht er schon bald die Kaufhausvitrine gegen die Bühne — ein Schauplatz, der es ihm erlaubt, seine Figuren, die in ihm schlummern, zum Leben zu erwecken. Objekte werden lebendig und fliegen ihm um die Ohren, Magie erfüllt den Raum, in welchem sich die Grenzen zwischen Fiktion und Realität verwischen. Bitte ändern Sie die Konfiguration Ihres Browsers. Yvonne Kummer Buchhaltung, Personal: Teil des Gesprächskonzertes mit Prof. Finden Sie gespeicherte Artikel schnell und einfach.
You can sit in the house We visited this place in July - i was about 36 degress in Wienna, really hot. This was nice warm evening. The restaurant is situated near wineyards on the hill so you have lovely views sitting outside.
We were really impressed. Great ambience in the garden during the summer months. Will be back next year. We have had dinner here in the garden. Great food and wine.
Asparagus was in season and where absolutely great! Best is to make a reservation. All of your saved places can be found here in My Trips.
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Could this location be considered a specialty food market? Is this a German restaurant? Is this a Central European restaurant? Can a vegetarian person get a good meal at this restaurant?
Is this restaurant appropriate for Kids? Is this a wine bar? Is this a place where you pay before receiving your order? Is this restaurant a hidden gem or off-the-beaten path?
Does this restaurant offer takeout or food to go? Share another experience before you go. Write a Review Reviews See what travelers are saying: Reviewed October 5, Great Schnitzel - careful how much you order!
Washington DC, District of Columbia. Reviewed August 10, via mobile. Great food, lovely atmosphere, not touristy. Reviewed April 5, Reviewed October 2, via mobile.
My favourite meal in Vienna! Reviewed July 3, Since some years, there is a worldwide trend to move towards ''higher-fidelity'' simulation techniques in reactor analysis.
One of the main objectives of the research in this area is to enhance the prediction capability of the computations used for safety demonstration of the current LWR nuclear power plants through the dynamic 3D coupling of the codes simulating the different physics of the problem into a common multi-physics simulation scheme.
In this context, the NURESAFE European project aims at delivering to the European stakeholders an advanced and reliable software capacity usable for safety analysis needs of present and future LWR reactors and developing a high level of expertise in Europe in the proper use of the most recent simulation tools including uncertainty assessment to quantify the margins toward feared phenomena occurring during an accident.
These physics are fully integrated into the platform in order to provide a standardized state-of-the-art code system to support safety analysis of current and evolving LWRs.
Large-scale irradiated fuel experiments at proteus research program. The techniques are based on delayed neutrons or high energy gamma-rays emitted by short lived fission products, both produced in pins re-irradiated in PROTEUS.
Preliminary results show that a 6x6 burnt fuel bundle should adequately represent the interface between fresh and burnt PWR fuel assemblies, and that CASMO-4E appears suitable to analyze this complex experiment.
Specifically, the analysis of the governing processes for the fuel rod behaviour during the RIA events simulated in the experimental facility of the Nuclear Safety Research Reactor NSRR, Japan are in the focus of the present study.
The results obtained can be useful for a better transfer of the NSRR test results in relation to the corresponding behaviour in LWRs and furthermore might also support the planning of future additional experiments.
For the development of new acceptance criteria for the analysis of Rod-Ejection-Accidents REA in Pressurized Water Reactor UO2 cores, full-width-at-half-maximum power pulse widths in the range 25—40ms were employed in the analytical transient fuel behaviour studies.
This study is presented in a first part of this paper. Finally, to provide confidence in the results, particularly for pulse widths below 25ms, sensitivity studies are performed to assess the effects of specific modelling options and assumptions applied in the REA 3D kinetic analysis and to compare the sensitivities between UO2 and MOX cores.
On the basis of all these studies, it is found that the MOX core pulse width is usually around 5 to 10ms lower than for UO2 cores.
For rod worths close to 2. To shortly address the applicability of these results, i. And it was found that even for these types of cores, the pulse width would, although being slightly reduced, remain in the range 10—15ms for rod reactivities up to pcm above prompt criticality.
In subsequent steps, radial and axial correction factors, both based on fast flux results from the 3-D core simulator, are applied to the calculated 2-D transport flux in order to take into account radial leakage effects as well as axial flux gradients around the tip.
The fluence is finally estimated through a time-integration of the corrected 2-D transport fast flux.
The developed methodology has been applied to estimate the fluence for a total of 15 control rods, over 21 operating cycles of a Swiss nuclear power plant.
An analysis of the influence of the recent upgrade in the data file is thus needed with respect to specific practical applications. A new LOCA test at Halden was planned as the first experiment within the Halden LOCA program addressing the behaviour of commercially irradiated BWR fuel of medium burn-up with burst of the cladding expected to occur at a temperature of about oC, which is essentially higher than in the preceding experiments.
The specific measures to be adopted have been suggested based upon a parametric study using the FALCON fuel behaviour code and aimed at an optimized design of the test fuel rod for the given high target cladding temperature of oC peak local.
The analysis has shown a reasonable agreement with the fundamental experimental findings, such as correlations of NUREG, as well as consistency with the data from Halden LOCA testing available so far.
Thus, a general conclusion is drawn about the applicability of the methodology developed at PSI to the analysis of LWR fuel rod behaviour during LOCA, in consideration of the effects of fuel burn-up.
Consistent evaluation of modern nuclear data libraries for criticality safety analyses of thermal compound low-enriched-uranium systems. For this purpose, validation calculations were performed for a suite of benchmarks from the International Handbook of Evaluated Criticality Safety Benchmark Experiments.
In order to define the ranges of applicability of the used calculational methods and to detect possible trends, the spectrum-related characteristics of the modeled critical experimental configurations were analysed.
All libraries have in common that any evidence of a statistically significant trend in the normalized eigenvalues versus both spectrum-related characteristics and experimental design parameters could not be found.
The mechanisms able to result in as significant fission gas release as measured by the post-test-examination are analysed with the model and discussed in consideration of the predicted initial micro-structural state of the fuel after the base irradiation.
The minor role of gaseous swelling in the early failure of the LS-1 test-fuel-rod is shown by calculation, which is due to premature brittle cracking of the highly hydrated cladding under the conditions of the Room-Temperature capsule used in the LS-1 test.
However, a significant potential impact of the gaseous swelling on cladding strain-stress conditions during the RIA is shown by the calculation when assuming a sufficient residual cladding ductility in the hypothetical test with the same parameters as in LS-1, but under conditions of the High-Temperature-High-Pressure capsule.
On the use of importance factors from Monte Carlo calculations for efficient fast neutron fluence prediction. Towards the development of upper subcriticality limits on the basis of benchmark criticality calculations.
This paper concerns the assessment of standard point-wise neutron data libraries for criticality safety evaluations in units of the effective neutron multiplication factor, keff, the aim being to establish a methodology for the analysis of storage pools containing fuel assemblies discharged from the Swiss Light Water Reactors.
Validation of standard neutron data libraries for LWR storage pools and transport casks criticality safety evaluations.
The benchmarks were selected on the basis of their similarity to designs found in today's Light Water Reactor LWR compact storage pools and transport casks, including MOX fuel rod assemblies, and in total comprised as much as cases.
It is found that the recently released JEFF On the effects of oxygen cross-sections in the fast neutron fluence analysis for a reactor pressure vessel scraping test.
Studies in support of the assessment of aging structural materials in pressurized water reactors are being performed at the Paul Scherrer Institut.
In the frame of the methodology validation, an investigation is currently reported pertaining to the sensitivity of the calculated results, for a specific reactor pressure vessel scraping test, to the nuclear data used with the Monte Carlo code.
Subsequent analysis has indicated that the observed discrepancy can be attributed mostly to differences in the oxygen data. The discussed cross-section differences could potentially lead to more sizable discrepancies for other applications, and thus need to be brought to the attention of neutron data evaluators and users.
The analytical capabilities established in that framework aim at providing a broad range of expertise related to the safety evaluation of the Swiss nuclear power plants NPPs.
A key requirement is that the employed code system provides the capability to perform at any given time, transient and safety analyses of the plants using as basis, validated 3-D core models up to the latest completed operating cycle.
Accordingly, the core analysis models need to be updated and re-qualified after each cycle outage, noting that eventual advances in terms of codes, methods and libraries need also to be integrated in this continuous process.
The objective of this paper is to present the core management system CMSYS that is being developed within STARS to ensure that the 3-D core modelling and analyses can be made in an efficient and consistent manner for all the Swiss plants.
The overall system architecture as well as the level of accuracy achieved for some of the Swiss 3-D core models is presented in a first part of the paper.
Thereafter, the advantages that the system offers in terms of assessing and qualifying new analysis methods are illustrated. The implementation of a simplified approach for efficient fast neutron fluence estimations is discussed in the paper.
The analysis is based on the utilisation of neutron importance factors obtained by means of standard forward neutron transport calculations with a Monte Carlo code.
The approach is supposed to be useful for a number of practical problems when precise calculations are either not necessary or not feasible.
Verification studies were performed based on comparison with results from detailed routine calculations that were validated against experimental data.
The results obtained indicate that the simplified approach yields rather satisfactory efficiency in terms of accuracy and computational expenditures, despite the fact that during the analysed reactor history the transition from an 'out-in-in' to a 'low-leakage' fuel management scheme caused a significant disturbance of the neutron source distribution.
Thus, provided that the reactor power and burnup distributions are known or reasonably well predicted, the proposed approach may be applied for the effective prediction of the fast neutron flux response and examples of some practical cases requiring such approximate techniques are also outlined.
A methodology is presented for the accurate assessment of the fast neutron fluence at the reactor pressure vessel RPV of a pressurised water reactor PWR.
The modelling is supported by sensitivity and optimisation studies, using precise MCNPX calculations and detailed reactor condition specifications.
As validation of the new methodology, an analysis has been performed to compare calculation results with values based on experimental data related to the RPV of a PWR.
The reference fluence estimates are based on a previously performed activity analysis of RPV scraping samples, taken after 10 cycles of operation of the nuclear power plant.
The agreement between computed results, i. This agreement is considerably better than that reported for the results of earlier calculations, based on use of the deterministic code BOXER.
In this paper, we present an uncertainty methodology based on a statistical approach, for assessing uncertainties in lattice code predictions of fuel composition changes with burnup due to uncertainties in the fuel geometric configuration, initial enrichment, and depletion conditions.
The methodology has been applied to depletion calculations with CASMO-4 and experimental data from the ARIANE Programme to estimate the calculation uncertainties in nuclide concentration and other neutronic parameters at any time during the irradiation history.
Results have shown that important information on the quality of the code's predictions can be obtained by analyzing the comparison of the code's estimates and their associated uncertainties, in the form of tolerance intervals, with experimental data and their reported errors.
Modeling the effects of axial fuel relocation in the IFA A model that investigates the thermal effects of the axial fuel relocation in high burn-up fuel rods during loss of coolant accidents LOCA is presented and applied to the analysis of IFA Equations are introduced for axial fuel mass conservation and radiation heat transfer which predict the time evolution of both the axial power profile and the specific linear heat stored by the fuel during the LOCA.
The model is based on the assumption that the cladding balloon can accommodate the fuel mass relocated or slumped from the upper fuel stack and is applied to parametrically investigate key relocation factors, such as the relocation length and the density of the relocated fuel in the balloon volume.
Calculations of the cladding and heater surface temperature histories are performed and compared with measurements. The model predictions are in good agreement with the thermocouple signals if the fuel mass density in the balloon volume is assumed to be relatively low.
Since PIE analyses have evidenced a large reduction of the fuel stack length due to axial relocation above the balloon location, our results could indicate the possibility of fuel mass dispersion from the rod after burst.
Spatial and model-order based reactor signal analysis methodology for BWR core stability evaluation. A new methodology for the boiling water reactor core stability evaluation from measured noise signals has been recently developed and adopted at the Paul Scherrer Institut PSI.
This methodology consists in a general reactor noise analysis where as much as possible information recorded during the tests is investigated prior to determining core representative stability parameters, i.
A central part in this approach is that the evaluation of the core stability parameters is performed not only for a few but for ALL recorded neutron flux signals, allowing thereby the assessment of signal-related uncertainties.
In addition, for each signal, three different model-order optimization methods are systematically employed to take into account the sensitivity upon the model-order.
The current methodology is then applied to the evaluation of the core stability measurements performed at the Leibstadt NPP, Swit-zerland, during cycles 10, 13 and The results show that as the core becomes very stable, the method-related uncertainty becomes the major contributor to the overall uncertainty range while for intermediate DR values, the signal-related uncertainty becomes dominant.
However, as the core stability deteriorates, the method-related and signal-related spreads have similar contributions to the overall uncer-tainty, and both are found to be small.
The PSI methodology identifies the origin of the different contributions to the uncertainty. Fur-thermore, in order to assess the results obtained with the current methodology, a comparative study is for completeness carried out with respect to results from previously developed and applied procedures.
The results show a good agreement between the current method and the other methods. Simulation of the Halden IFA At the Paul Scherrer Institut PSI , a computational methodology was developed to model the thermal-hydraulics and thermo-mechanical behaviour of coolant channels and rods used in the LOCA tests performed at the Halden rector in Norway.
Results show good agreement between TRACE calculations of cladding temperature histories and measurements.
A preliminary application of the methodology to the IFA Such relocation is simply modelled by suppressing the impact of the fuel heat source after the onset of burst.
Analysis of the balloon size calculated by FALCON indicates a possible loss of fuel column between 5 and 17 cm, in reasonable agreement with preliminary observations at Halden.
First, the original FALCON code version MOD01 is modified by introducing new routines yielding a better description of the initial fuel and clad geometry, which allows the modelling of highly oxidized claddings with non-uniform axial oxide thicknesses.
By calculating the Hoop strain exerted by the oxide layer on the cladding, the correct initial relative position between inner and fuel clad outer surfaces is obtained.
Application of the new algorithm to the REP-NA4 experiment and comparison with the previous code results shows that the new version yields better predictions of the final clad outer diameter profile.
The new code version is then employed to predict the thermal-mechanical behaviour of the CIP test rod. The measured clad diametral deformation, clad elongation and coolant temperature are taken as figures of merit, the code calculations being compared against these.