Tamas Janos Katona | University of Pecs (original) (raw)

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Papers by Tamas Janos Katona

Földrengés után egy atomerőműben tudni kell, hogy az alapvető biztonsági funkciók megvalósításáho... more Földrengés után egy atomerőműben tudni kell, hogy az alapvető biztonsági funkciók megvalósításához szükséges rendszerek milyen állapotban lehetnek, ez szabja meg az üzemzavar elhárítás, illetve a földrengést követően a visszaindítás feltételeit. A pusztító hatást jellemezheti a maximális vízszintes talajgyorsulás, a válaszspektrum és a különböző műszeresen mérhető intenzitás-értékek. Ezek általában szelektívek egy-egy károsodási típusra, egyesek jól korreláltathatók úgy a rengés méretével, távolságával, mint a szerkezet-típusok viselkedésével. A cikkben megvizsgáljuk a földrengés jellemzők és a károsodási mechanizmusok közötti összefüggést, s elemezzük a földrengés jellemzők kár-indikátorként való alkalmazását a földrengés utáni visszaindíthatóság kritériumaként, illetve az állapot-értékelésre való alkalmasság és a felülvizsgálatok tervezése szempontjából.

The fundamental features of thermohydraulic fluctuations characterising the subcooled boiling in ... more The fundamental features of thermohydraulic fluctuations characterising the subcooled boiling in the PWR core have been investigated with the aim of boiling detection via neutron noise analysis. The relationship between different formulations of the model for thermohydraulic fluctuations and the inportant contributing phenomena has been exanine6 along the time scale of interest. A quasi-adiabatic one-channel coolant nodel has been derived for the description of fluctuations of flow variables. ±he effects of thermal nonequilibriuF feature of the two-phase flow on the ther~nydraulic fluctuations and the relationship between their statistical characteristics and engineering parameters cf the flow have been analysed. KEYIqORDS PWR, subcooled boiling, neutron noise, nonequilibriun~ two-phase flow, therzohy~raulic fluctuations, one-channel n odel.

ABSTRACT Method is worked out for the calculation of the acoustic eigenfrequencies, amplitude dis... more ABSTRACT Method is worked out for the calculation of the acoustic eigenfrequencies, amplitude distributions, APSD and CPSD values of pressure fluctuations in the primary circuit, taking into consideration also the friction. The method is based on an analogy of the passive quadripol characteristics and the acoustic fluctuation transfer matrix behaviour.The acoustic eigenfrequencies of the PWR primary circuit at different operating modes were calculated. The resonant frequencies were related to the different primary components and their different thermohydraulic conditions, respectively.

ABSTRACT A modified theory of propagating temperature fluctuations is developed, in which the noi... more ABSTRACT A modified theory of propagating temperature fluctuations is developed, in which the noise source is inlet temperature fluctuation, a fluctuation of coolant velocity or the fluctuation of reactor power. Void fluctuations in the upper part of the channel are induced by the varying degree of subcooling due to fluctuations of the liquid temperature. The theoretical expressions are evaluated numerically and the results are compared with PWR measurements.

Understanding and Mitigating Ageing in Nuclear Power Plants, 2010

Progress in Nuclear Energy, 1988

ABSTRACT In the paper the characteristic sinks in the Cross Power Spectral Density (CPSD) and coh... more ABSTRACT In the paper the characteristic sinks in the Cross Power Spectral Density (CPSD) and coherency functions of neutron and temperature noise in PWRs are studied. The experimental results and theoretical models for the interpretation of the sink structure are discussed from the point of view of coolant velocity and subcooled boiling determination. A one-dimensional coupled thermohydraulic neutronic reactor noise model is applied for the better understanding of the sink - structure of the characteristic functions of the neutron and temperature noise in PWRs.

Nuclear Engineering and Design, 1998

The dynamic response of structures due to seismic loadings is conventionally analyzed in the time... more The dynamic response of structures due to seismic loadings is conventionally analyzed in the time domain using substructure methods (decoupled system models). This procedure uses frequency-independent impedances to represent capabilities of the soil underneath the structure. The soil parameters are tuned to the fundamental frequencies of the soil-structure system. This is a common procedure widely used in the preliminary design of power plant structures which provides conservative results. However, parallel to the rapid progress being made in upgrading the capability of data processing systems, methods and software tools have become available which work also in the frequency domain using complex models (for the soil and the structure) or models in which the soil is represented by frequency-dependent impedances. This procedure (coupled system models) also allows realistic treatment of kinematic interaction effects and especially consideration of the embedment parameters of the building structure. The main goal of the study presented here was to demonstrate the effects of different procedures for consideration of soil-structure interaction on the dynamic response of the structures mentioned above. The analyses were based on appropriate mathematical models of the coupled vibrating structures (reactor building, turbine hall, intermediate building structures of a VVER 440/213 as well as the main building of the VVER 1000) and the layered soil. On the basis of this study, it can be concluded that substructure methods using frequency-independent impedances (equivalent dashpots) and cut-off of modal damping usually provide conservative results. Coupled system models which allow the soil-structure interaction effects to be realistically represented (by coupled models of the soil and the structure or by frequency-dependent impedances) provide more accurate results. The advantage of the analysis using coupled system models will be demonstrated and discussed, based on results obtained for the VVER 440/213 PAKS and VVER 1000 Kozloduy.

The design life of the NPP PAKS WWER-440/213 units is 30 years. Formally the design life limits t... more The design life of the NPP PAKS WWER-440/213 units is 30 years. Formally the design life limits the operational licence. The condition of the equipment after almost 20 years of operation is excellent. The lifetime management activity of the Paks NPP provides a proper basis for the continuation of safe operation. The current Hungarian Nuclear Regulations allow the extension of the operation licence of the NPP Units beyond the design life. To prove the acceptability of continuation of operation the Licensee have to prepare a Licence Renewal Application not later than 5 years before the expiration of the design lifetime. The Licence Renewal process based mainly on USNRC approach taking into account the special features of the WWER-440/213 Units. Summary of the preconditions of the Licence Renewal and the activity for the preparation of the operational lifetime extension are reported in the paper.

Magyar tudomány: [a Magyar Tudományos Akadémia Értesítője]

ABSTRACT The strategic goal of the VVER operator is to extend its operational lifetime beyond the... more ABSTRACT The strategic goal of the VVER operator is to extend its operational lifetime beyond the design life. Here, technical and regulatory conditions and methods for ensuring long-term operation of the VVER plant are presented plus an overview of the basic technical design features of VVER relevant to long-term operation. Degradation mechanisms of structures and components which limit the operational lifetime of the plants are identified. The method for evaluating ageing of the plant, a review of existing plant activities for ensuring the required performance of safety-related systems, development of ageing management programmes and other related plant programmes are described. The integration of plant programmes into a system that ensures safe long-term operation is shown through examples. Trends and need for future research are presented.

Pollack Periodica, 2011

Assessment of seismic safety of nuclear power plants requires convolution of plant fragilities wi... more Assessment of seismic safety of nuclear power plants requires convolution of plant fragilities with seismic hazard curves. Plant level fragility versus peak ground acceleration is obtained by combining component fragilities according to the Boolean-expression of the sequence leading to core damage. An improvement of the fragility modeling can be achieved via description of the fragility as a function of cumulative absolute velocity of ground motion instead of peak ground acceleration. In the paper the physical meaning and dependence on strong motion parameters is also discussed. The reason why the cumulative absolute velocity is an appropriate damage indicator is analyzed.

Volume 7: Operations, Applications and Components, 2009

Volume 7: Operations, Applications and Components, 2012

One of the most complex cases for assessing the nuclear power plant safety is the evaluation of t... more One of the most complex cases for assessing the nuclear power plant safety is the evaluation of the response of the plant to an earthquake and calculation of the core damage frequency related with this. Plant level fragilities are convolved with the seismic hazard curves to obtain a set of doublets for the plant damage state. The standard methodology of the description of randomness and epistemic uncertainty of the fragility is based on the use of lognormal distribution. In the practice, because of large number and variety of types of components, variety of failure modes, further simplification is needed in spite of simplicity of the mathematic description of the fragility and its uncertainty. Sophisticated modeling and screening methods have to be applied for plant fragility development requiring enormous experience. Several practical assumptions utilized in the seismic PSA showing certain analogy with interval type description of uncertainties. In the paper an attempt is made for outlining some new options for nuclear power plant seismic fragility development based on the interval and p-box concept. The possibility for derivation of conditional probability of failure for cumulative absolute velocity is also highlighted.

The Paks Nuclear Power Plant is constructed at the soft soil site. As a result of the post-Fukush... more The Paks Nuclear Power Plant is constructed at the soft soil site. As a result of the post-Fukushima stress-test, liquefaction should be analysed as beyond design basis hazard. The aim of the analysis is to define the post-event condition of the plant, definition of plant vulnerabilities and identification of the necessary measures for accident management. In the paper, the methodology of the analysis of liquefaction effects for nuclear power plants is outlined. The procedure includes identification of the scope of the safety analysis and the acceptable limit cases for plant structures having different role from accident management point of view. Considerations are made for identification of dominating effects of liquefaction. The possibility of the decoupling of the analysis of liquefaction effects from the analysis of vibratory ground motion is discussed. It is shown in the paper, that the practicable empirical methods for definition of liquefaction susceptibility provide rather controversial results. Selection of method for assessment of soil behaviour that affects the integrity of structures requires specific considerations. The case of Nuclear Power Plant at Paks, Hungary is used as an example for demonstration of practical importance of the presented results and considerations.