Institute of Safety Research; Annual Report 1993

The report gives an overview on the scientific work of the Institute of Safety Research in 1993

Institute of Safety Research; Annual Report 1997

The report gives an overview on the scientific work of the Institute of Safety Research in 1997

Institute of Safety Research, Annual Report 1994

The report gives an overview on the scientific work of the Institute of Safety Research in 1994

Institute of Safety Research, Annual Report 1995

The report gives an overview on the scientific work of the Institute of Safety Research in 1995

Aktuelle Themen der Reaktorsicherheitsforschung in Deutschland

Die Veranstaltung widmete sich mit der Borverdünnung in Druckwasserreaktoren bzw. mit der Verstopfung der Sumpfansaugsiebe durch freigesetztes Isolationsmaterial schwerpunktmäßig zwei Themen der Reaktorsicherheit, die auch in aktuellen Aufsichtsverfahren eine Rolle spielen. Eingebettet in den internationalen Kontext wollten die Veranstalter die sicherheitstechnische Bedeutung dieser Themen für die deutschen Anlagen beleuchten und die Auswirkungen auf die zu erbringenden Sicherheitsnachweise und den Anlagenbetrieb darstellen. Dabei kamen Gutachter, Vertreter der Forschung, Hersteller und Betreiber gleichermaßen zu Wort. Der Fachtag sollte den Teilnehmern aber insbesondere vermitteln, welche Beiträge die privat und öffentlich finanzierte Reaktorsicherheitsforschung zur Aufklärung der jeweiligen Ereignisabläufe und ihrer sicherheitstechnischen Bedeutung geleistet hat. In diesem Forschungskontext spielen, auch international, die Methoden der so genannten Computational Fluid Dynamics (CFD) eine zunehmende Rolle. Deshalb widmete sich eine Sitzung den Grundlagen, Möglichkeiten und Grenzen von CFD-Methoden. Dabei wurden u.a. Anwendungen zur Borvermischung und zum Verhalten von Mineralwolle im Sumpf präsentiert

Annual report 2000 Institute of Safety Research

The report gives an overview on the activities of the Institute of Safety Research in 2000

Acoustic Leak Detection at Complicated Geometrical Structures Using Fuzzy Logic and Neural Networks

Methods of acoustic leak monitoring are of great practical interest for the safety of pressure vessels and pipe lines not only at the primary circuit of nuclear power plants. In this report some aspects of acoustic leak localization at complicated three-dimensional topologies for the case of leakage monitoring at the reactor vessel head of a VVER-440 are discussed. An acoustic method based on pattern recognition is being developed. During the learning phase, the localization classifier is trained with sound patterns that are generated with simulated leaks at all locations endangered by leak. After training unknown leak positions can be recognized through comparison with the training patterns. The sound patterns of the simulated leaks are simultaneously detected with an AE-sensor array and with high frequency microphones measuring structureborne sound and airborne noise, respectively. The initial results show the used classifiers principally to be capable of detecting and locating leaks, but they also show that further investigations are necessary to develop a reliable method

ICONE 14-89120 BUOYANCY DRIVEN COOLANT MIXING STUDIES OF NATURAL CIRCULATION FLOWS AT THE ROCOM TEST FACILITY USING ANSYS CFX

ABSTRACT Coolant mixing in the cold leg, downcomer and the lower plenum of pressurized water reactors is an important phenomenon mitigating the reactivity insertion into the core. Therefore, mixing of the de-borated slugs with the ambient coolant in the reactor pressure vessel was investigated at the four loop 1:5 scaled ROCOM mixing test facility. Thermal hydraulics analyses showed, that weakly borated condensate can accumulate in particular in the pump loop seal of those loops, which do not receive safety injection. After refilling of the primary circuit, natural circulation in the stagnant loops can re-establish simultaneously and the de-borated slugs are shifted towards the reactor pressure vessel (RPV). In the ROCOM experiments, the length of the flow ramp and the initial density difference between the slugs and the ambient coolant was varied. From the test matrix experiments with 0 resp. 2% density difference between the de-borated slugs and the ambient coolant were used to validate the CFD software ANSYS CFX. To model the effects of turbulence on the mean flow a higher order Reynolds stress turbulence model was employed and a mesh consisting of 6.4 million hybrid elements was utilized. Only the experiments and CFD calculations with modeled density differences show a stratification in the downcomer. Depending on the degree of density differences the less dense slugs flow around the core barrel at the top of the downcomer. At the opposite side the lower borated coolant is entrained by the colder safety injection water and transported to the core. The validation proves that ANSYS CFX is able to simulate appropriately the flow field and mixing effects of coolant with different densities

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)

Ein technisches Informationssystem zur verbesserten betrieblichen Überwachung des Kernkraftwerkes Saporoshje/Ukraine

In order to improve the operational surveillance of a VVER-1000 unit of the Ukrainian nuclear power plant Zaporosh´ye a technical monitoring system has been specified. The system will enable the state regulatory and supervisory bodies to survey the unit operation independently of operators to assess its safety status, and to impose appropriate conditions. Due to its up-to-date configuration the system provides early indication of any operational incident and emission of radioactive materials connected. Based on the system an immediate warning in mergency situations is possible as well as an effective emergency management. For this purpose 49 different operational parameters of the unit, 18 radiological parameters of the unit and the plant site and 6 meteorological parameters are monitored. The monitoring concept and its technical realization are described