[en] The operation experience for the BWR and PWR type nuclear power reactor fuel elements is generalized. Modern aspects of operational reliability of light water cooled reactor fuel elements, in particular the problem of decreasing the fuel element failure probability are discussed. Economic aspects of fuel element improvement and their influence on the cost of electric energy produced at nuclear power plants are analyzed. A sufficiently high operational reliability of light water cooled reactor fuel elements is pointed out. The part of depressurized fuel elements in a core per year of reactor operation constitutes 0.05%. The improvement of the fuel element design and technology as well as creating more strict operational conditions eliminated fuel element failures due to deposit formation hydrogenation and can breaking. Most dangerous nowadays is fuel element failure because of mechanical fuel-cladding interaction. The conclusion is drawn that perfection of the design and technology of fuel element manufacturing improvement of the fuel utilization as well as changes in NPP operation conditions make it possible to obtain a considerable economic gain

[en] The influence of strain rate on yield point of alloy Zr-1%Nb was under study. Zirconium alloy tube specimens were loaded with pulsed internal pressure. Within the strain rate range 10⁰ to 10³ s^-1 the yield strength varies from 500 to 570 MPa. The values of yield strength observed for tube specimens are pointed out to be higher compared to those for rod-like ones. This is related to the difference of their textures. The experiments were accomplished using zirconium alloy tubes intended for manufacturing fuel cans

[en] A method and results of determination of conventional yield strength of the alloy Zr-1%Nb of the thin-walled tube around it upon loading the tube with pulse internal pressure are presented. The experiments are performed on a magnetopulse apparatus at tree deformation rates in the range of 1-10³ sec^-1. The yield strength of the tube alloy in the specified range increases from 500 to 570 MPa. It is found that the yield strength of the tube alloy exceeds that of the cylinder rod specimens by 40-455 MPa, which is attributed to the difference in the alloy structure

[en] In many respects efficiency of measures taken to eliminate accident conditions at NPP with PWR that can occur due to loss of coolant by the primary circuit is determined by the conditions of filling the core with a coolant from the emergency cooling system. It is therefore important that during an accident the ballooning of fuel clads should not significantly block the flow area of fuel assemblies or marginally impair their cooling. Thus, the assessment of the extent and character of LOCA induced blockage of a flow area in water reactor assemblies is an important component in the analysis of an NPP safety. This report discusses the probability model and procedure of assessing the blockage of a flow area of VVER-type assemblies in LOCA

[en] Following an accidental loss of primary coolant (LOCA) in a Pressurized Water Reactor, there is the possibility that 'ballooning' or swelling of the fuel element cladding may occur, impairing the cooling produced when the reactor is reflooded with emergency cooling water. A computer program has therefore been developed with which to calculate the magnitude of this impairment. It is shown that, in general, it will be insignificant; in only 4% or less of the sub-channels between fuel elements will the blockage due to clad-ballooning be of any importance. (author)

No abstract available

No abstract available

[en] RAPTA-5 code used for licensing calculations to validate the compliance with the requirements for WWER fuel safety in design basis accidents. The characteristic results are given of design modelling experiments simulating thermomechanical and corrosion behaviour of WWER and PWR fuel rods in LOCA. The results corroborate the adequate predictability of both individual design models and the code as a whole. (author). 14 refs, 12 figs

[en] A brief description of RAPTA-SFD code intended for computer simulations of WWER-type fuel elements (simulator or absorber element) in conditions of accident with severe damage of fuel. Presented are models of chemical interactions of basic materials of the active zone, emphasized are special feature of their application in carrying out of the CORA-W2 experiment within the framework of International Standard Problem ISP-36. Results obtained confirm expediency of phenomenological models application. (author). 6 refs, 7 figs, 1 tab

[en] Experimental data on mechanical properties by Zr - 1% Nb alloy strain within the deformation rate range of 2.5 x 10^-3 - 8 x 10⁴ s^-1 and temperature range of 20-700 deg C are presented. Increase in the deformation rate within the above range leads to growth in strength characteristics and reduction of the alloy plasticity. The dynamic load (3000 s^-1) suppresses the effects of dynamic deformation ageing and the alloy superplasticity, observed by static deformation rate