[en] An analysis of reliability of the WWER type cooling system pumps was carried out. It estimated the probability of a different type of cooling system failure. The analysis of the structure reliability was carried out by the fault tree method. The analysis of the thermal reliability of the reactor core was carried out in these conditions. 3 refs

[en] A program PODER for processing data from the operating modes of the reactors taking into account the effects of corrosion, hydration, and deformation of the nuclear reactor fuel element sheathing, the formation of the corrosion product deposits, the change in the geometric dimensions of the nuclear reactor fuel element due to the temperature deformation, as well as the various gas fillers, are elaborated. The ''hot channel'' method determining the reliability of the system is realized. The basic equations describing the thermohydraulic processes in nuclear reactors are solved by the finite difference method. Approximations are carried out with the approach of least squares. The temperature distribution versus the zirconium sheathing height is computed for the case of WWER-440 type reactors. The advantages of the proposed program P0DER are discussed

[en] Results from reactor core calculations with the RELAP4/MODE6 code for the blowdown phase of loss-of-coolant accident (LOCA) in WWER-440 reactors are presented. The core boundary conditions are obtained from a previous primary loop calculation using the same code. An analysis is given of the thermohydraulic parameters in channels with different hot channel factors, and the thermomechanical behaviour of fuel rods is evaluated. A conclusion about reducing the conservatism of the results is drawn

[en] In connection with the forthcoming construction of a npp with the wwer-1000 reactor the loss of coolant accident associated with the main circulation tube rupture at the inlet near the reactor is analyzed. The relap4/mod6 program is used for the analysis. The data obtained show that the coolant outflow stage continues for about 25s. On the average the pressure in the circuits varies from 16 to 10 mpa per 0.1s and then it continues to decrease slowly. The pressure in the steam generator at the secondary circuits end increases approximately up to 6.9 MPa as a result of steam generator blocking and remaining coolant heating and then somewhat decreases owing to the primary circuit cooling. By the end of the fuel and can temperatures are equalized and the heat transfer coefficient is stabilized at the level of 100 w/1 (m²xK). It is concluded that during a loss of coolant accident at the wwer-1000 reactor in procesess of coolant blowdown in the medium power fuel elemets neither the fuel, melting temperature (3000 k), nor the critical temperature (1000 k) of plastic deformation zirconiu can initiation are attained

[en] A heat-transfer model during the formation of corrosion product deposition on the fuel element cladding is proposed. A method of calculating the heat quantity required for the evaporation of the coolant on the surface of the nuclear reactor fuel element is proposed which characterizes the quantity of the formed precipitates. A good agreement of the results obtained by this method and the experimental data exists. A relation of the effective heat conductivity of the corrosion product deposition is obtained in analytical form

[en] For the first time the analysis results of the processes in fuel elements are presented for all three MDA stages (break flow, lower plenum filling, core reflooding from the emergency cooling system), considered time interval is O-200s. The impact of some major initial parameters on the fuel rod cladding temperature and strains during MDA has been investigated. To define the initial conditions including the state of the primary circuit and reactor core RELAP 4/Mod 6 codes were used. The processes in the fuel rods were calculated by SSYST-2 code combined with the first two into a single complex. The following factors were investigated: fuel power density, initial gas pressure under the cladding, design tolerances, fuel burnup. It is pointed out that the described code complex is suitable for the thermomechanical analysis of the WWER-1000 fuel rods and provides reliable results

[en] The RELAP4/MOD6 code was used to analyse the thermal hydraulics of the primary circuit and the core of a WWER-440 reactor during the blowdown phase of a loss-of-coolant accident. The influence of the accumulators on the blowdown duration and the fuel rod surface temperature was evaluated. A parametric study of the hot spot factor influencing the cladding temperature was carried out

[en] A calculation model was developed for evaluating of reactor parameter random deviations of the fuel rod surface temperature during large LOCA of a WWER-type reactor. The model proposed is based on the widespread methods for one dimensional fluid calculations detailed by S. Fisher in RELAP 4/MOD6. An analysis of the separate contributors in fuel rod surface temperature standard deviation during the aacident shows that their effect depends on the chosen time interval. The method proposed can also be used for calculating the rest reactor parameter deviations during LOCA

[en] A computational model for evaluating the influence of random variations in some reactor parameters on the fuel rodsurface temperature during large LOCA for WWER type reactors is described. The proposed procedure is based on one of the widely used methods for one-dimensional thermal hydraulic computations and can be successfully used for sensitivity analysis of the LOCA model predictions

[en] The WWER-1000 system behaviour during a LOCA coincident with a mechanical failure of the main cooling pump is analysed using the codes RELAP-4/06 and SSIT-2. Data on the changes in the most important thermophysical parameters of the primary coolant cirquit, reactor core and the fuel elements under these conditions are obtained. The high reliability of the fuel elements of WWER-1000 type reactor during the above-mentioned regimes is proven. 6 figs., 4 refs