[en] This study is a part of the IAEA coordinated research program 'Benchmark study for the Seismic Analysis and Testing of VVER Type NPPs'. The study reports the numerical simulation of the blast test for Paks and Kozloduy nuclear power plants beginning from the recorded free-field response and computing the structural response at various points inside the reactor building. The full-scale blast tests of the Paks and Kozloduy NPPs took place in December 1994 and in July 1996. During the tests the plants operated normally. The instrumentation for the tests consisted of 52 recording channels with 200 Hz sampling rate. Detonating 100 kg charges in 50-meter deep boreholes at 2.5-km distance from the plant carried out the blast tests. The 3D structural models for both reactor buildings were analyzed in the frequency domain. The number of modes extracted in both cases was about 500 and the cut-off frequency was 25 Hz. In the response history run the responses of the selected points were evaluated. The input values for response history run were the three components of the excitation, which were transformed from time domain to the frequency domain with the aid of Fourier transform. The analysis was carried out in frequency domain and responses were transferred back to time domain with inverse Fourier transform. The Paks and Kozloduy blast tests produced a wealth of information on the behavior of the nuclear power plant structures excited by blast type loads containing also the low frequency wave train if albeit with small energy content. The comparison of measured and calculated results gave information about the suitability of the selected analysis approach for the investigated blast type loading

[en] The report summarises the studies carried out in Fortum Engineering (formerly IVO Power Engineering) between the years of 1992 and 1997 concerning ultimate strength of structures designed to mitigate and contain the consequences of various core melt accident scenarios. The report begins with the description of containment loading situations arising from core melt accidents. These situations are divided to fast and slow loads. Fast loads include ex-vessel steam explosions, steam spikes, hydrogen burns, direct containment heating and missiles. Slow loads are connected with pressure rise inside the containment in case when the containment heat removal system is not functioning. First part of report describes the analyses of reactor cavity based on axi-symmetric load assumptions. These studies are performed with various models like one degree of freedom idealisation, axi-symmetric modelling of geometry and full three-dimensional modelling of geometry. Second part of report describes the analyses of cavity based on non-axi-symmetric load assumptions. Here full 3D- geometry model is used combined with various physical models for the behaviour of reinforced concrete. Third part of report gives short account of the analysis of containment ultimate pressure capacity. The containment model in this case includes pre-stressing tendons and mild steel reinforcing bars. The load is assumed to axi-symmetric internal static pressure. The capacity of the reactor cavity against the ex-vessel steam explosion scenarios for VVER-91 plant concept is established for both axi-symmetric and non-axi-symmetric load models using ANACAP structural analysis code. The validation of the cavity response to ex-vessel steam explosion load using different commercially available codes gave mixed results for both axisymmetric and non-axi-symmetric load presentations.The ultimate static overpressure capacity of the VVER-91 reactor cavity structure was established to be of the order of 10 MPa. This result was convincingly validated by all used programs, namely, ANACAP, ANSYS and ABAQUS/STANDARD. This capacity value is applicable for steam spikes.The first bounding study for ultimate over pressure capacity for VVER-9 1 containment building was carried out. The capacity value obtained was 8 MPa. This value cannot be considered conclusive. (orig.)

[en] Seismic load is in many areas of the world the most important loading situation from the point of view of structural strength. Taking this into account it is understandable, that there has been a strong allocation of resources in the seismic analysis during the past ten years. In this study there are three areas of the center of gravity: (1) Random vibrations; (2) Soil-structure interaction and (3) The methods for determining structural response. The solution of random vibration problems is clarified with the aid of applications in this study and from the point of view of mathematical treatment and mathematical formulations it is deemed sufficient to give the relevant sources. In the soil-structure interaction analysis the focus has been the significance of frequency dependent impedance functions. As a result it was obtained, that the description of the soil with the aid of frequency dependent impedance functions decreases the structural response and it is thus always the preferred method when compared to more conservative analysis types. From the methods to determine the C structural response the following four were tested: (1) The time history method; (2) The complex frequency-response method; (3) Response spectrum method and (4) The equivalent static force method. The time history appeared to be the most accurate method and the complex frequency-response method did have the widest area of application. (orig.). (14 refs., 35 figs.)

[en] System identification allows to build mathematical models of a dynamic system based on measured data. System identification is carried out by adjusting parameters within a given model until its output coincides as well as possible with the measured output. The aim of this study is to investigate and model the behavior of complex vibratory systems on the basis of measured excitation and response. The first part of the study describes the theory used in the analysis and the software tools used in the analysis. The second part of the study describes the investigation and modeling of the response of single degree of freedom oscillator excited by sinusoidal and blast excitation. In the third part of the study the system identification of the Kozloduy NPP unit 5 reactor building and Paks NPP unit 1 reactor building is studied and the models are estimated using the method of segmentation of excitation and response. System identification is carried out using MATLAB software by adjusting parameters within a given model until its output coincides as well as possible with the measured output. The types of models used for the were: l) ARX models; 2) ARMAX model; 3) Output-Error (OE) models; 4) Box-Jenkins (BJ) models; 5) State-space models. The model coefficients for different models were calculated using the least-squares and maximum likelihood estimation methods available in MATLAB system identification toolbox. Excitation was in both Paks and Kozloduy case the measured free-field excitation and responses were the vibration responses of the building on the foundation slab level and top of the building. By examining the established models the frequency characteristics of vibration systems were determined with 95 % accuracy and the amplitude response with 80 % accuracy. In case of the steady state response of sinusoidally excited single dof oscillator the modelling gave almost exact results. But in the case of the blast response of the reactor building the obtaining of the accuracy reported in last paragraph required segmentation of the 5 second blast response to 0.5 s segments better results results could be obtained if the segmentation was carried out to the level of 0.1 s segments but this required excessive amount of work and even in this case the agreement was not perfect. In conclusion the applicability of the available software tools of system identification to complex vibratory systems give on satisfactory results even with great numerical effort applied. (orig.)

[en] There are many different important problem areas in evaluating the seismic response of structures. In this study the effort is concentrated on three of these areas. The first task is the mathematical formulation of earthquake excitation. The random vibration theory is taken as the tool in this task. The second area of interest in this study is the soil-structure interaction analysis. The approach of impedance functions is chosen and the focal point of interest is the significance of frequency dependent impedance functions. The third area of interest is the methods to determine the structural response. The following three methods were tested: the mode superposition time history method; the complex frequency response method; the response spectrum method. The comparison was made with the aid of MSC/NASTRAN code. The three methods gave for outer containment building response results which were in good agreement with each other. (author). 4 refs., 5 figs

[en] The failure probability assessment of the containment building is an essential feature of the Level 2 PSA-studies of nuclear power plants. The geometry of the containment was determined by the preliminary design. The seismic hazard of the plant site was assessed during Level 1 PSA of Loviisa plant. The initial information for seismic fragility analysis of the containment is the seismic response of the structure. The structural model for response analysis was the stick model. The stress analysis of the containment was carried out using the shell element model. The fragility evaluation of the containment was performed with the PROBAN-program. The structure was modeled as a parallel system consisting of the most heavily stressed elements. The resulting fragility curve gives the conditional probability of failure as a function of peak ground acceleration. The seismic hazard and the fragility were convolved to obtain the annual nonexceedance probability distribution for the collapse frequency of the structure. The dominating range of this distribution was of 1E-12 to 1E-8. (author)

[en] The purpose of the present work is the estimation of seismic hazard in the territory of the nuclear power plant OL3 in Olkiluoto. Because there are no registered strong motion acceleration recordings of earthquakes in Finland, the earthquake recordings from Saguenay and Newcastle regions from Canada and Australia were taken as sources of initial data because of their geological and tantalic similarity to Fennoscandia. The code basis for the ground motion estimation in probabilistic seismic hazard studies stipulates the median spectra for mean return period of 100 000 years. The decision three approach is used in the treatment of uncertainties in this study. The hazard analysis was carried out with the aid of Fortran Computer Program for Seismic Risk Analysis SEISRISK II developed by USGS. The resulting raw site hazard depicting the distribution of the seismic hazard was given in the form 32 curves equipped with appropriate weights. The end result of the analysis was given in the form of median, mean, 95% - fractile and 5% - fractile curves for horizontal peak ground acceleration amplitudes of 0.001 g, 0.005 g, 0.01 g, 0.05 g, 0.07 g, 0.1 g, 0.2 g, 0.3 g and 0.4 g and for corresponding non-exceedance probabilities. (author)

[en] The failure probability assessment of the containment building is an essential feature of the Level 2 PSA studies of nuclear power plants. The primary purpose of this paper is to demonstrate the methodology of evaluating containment seismic induced probability of failure without containment pressurization. The Loviisa, Finland site is one of the most seismically stable in the world and the numerically evaluated seismic induced failure probabilities are not representative for other sites. In addition, the containment concept described in this paper is not the typical Russian design which uses helical tendons in the cylindrical part of the structure and has a ring girder at the spring line of the structure. So the conclusions reached are applicable only to the containment configuration described in the paper. The geometry of the containment was determined by its preliminary design. The seismic hazard of the plant site was assessed during Level 1 PSA of the Loviisa plant. The initial information for seismic fragility analysis of the containment is the seismic response of the structure. The structural model for response analysis was the stick model. The stress analysis of the containment was carried out using the shell element model. The fragility evaluation of the containment was performed with the PROBAN-program. The structure was modeled as a parallel system consisting of the most heavily stressed elements. The resulting fragility curve gives the conditional probability of failure as a function of peak ground acceleration. The seismic hazard and the fragility were convolved to obtain the annual nonexceedance probability distribution for the collapse frequency of the structure. (orig.)

[en] The reactor cavity of VVER-91 NPP is a thick-walled, cylindrical reinforced concrete structure. In case of molten core-water reaction during the severe reactor accident the load carrying capacity of the cavity structure is of interest against the short impulse type loading caused by the steam explosion phenomenon. The assumed size of the impulse was 20 kPa-s and the duration was 10 ms. This investigation was divided in several phases. First, the elastic response of the cavity was determined using the ABAQUS code. Next, the static response of the cavity was evaluated using elasto-plastic properties of reinforcement and concrete and also taking into account the cracking of the concrete. This analysis was done with the aid of ABAQUS/STANDARD and ANSYS codes and the obtained results agreed reasonably with each other. In order to obtain a qualitative picture of the behaviour of the structure under the impulse load a simplified single degree of freedorn model was developed. The hoop reinforcement of the cavity was taken as an elasto-plastic spring and the wall concrete acted as a mass. Using this model the suitable amount of hoop reinforcement was determined. In next phase, the dynamic analysis of the structure was attempted using elasto-plastic material properties and concrete cracking. (13 refs., 57 figs.)

[en] The type of the containment building is a double-containment, the purpose of which is to produce the isolating protection between the processes of the reactor building and the environment. The outer concrete shell of the containment building gives the external protection to the process and to the inner steel shell against the future effects of the environment and the purpose of the inner steel shell is to prevent the emissions from getting directly into the environment in any process situation. The free-standing inner steel shell has been anchored from its bottom on the ring plate on the elevation +9.60 which from the middle part extends downwards as the reactor pit reaching the reactor pit base slab on the bedrock. These reinforced concrete structures together with the vertical structures, which support the material air lock and the elevation +9.60, form the entity which in this report is called the reinforced concrete part of the containment building. The loads used in the original design of the containment are given