[en] Parts and materials, having been stored for a long period of time (32 years) in the storage of IGR reactor used fuel, was examined. Uranium-graphite blocks, steel sealed cans with boron carbide powder, steel semi-hermetic tanks for fuel blocks storing, steel frameworks for tanks, leaden blocks, concrete cover on the bottom and wall of storage dry area, dry area protective lid concrete sections were visually examined. Steel parts surface corrosive damage depth was evaluated. Rather satisfactory state of all the parts and materials of the storage was determined on the basis of obtained data. (author)

[en] The Atomic Energy Institute of the National Nuclear Center of Republic of Kazakstan with cooperation with Russian research institutes and atomic energy enterprises were conducted experimental explorations for the validation atomic energy safety providing to IGR impulse reactor. Methodical experience that has been worked out in capsule testing and the complex of available equipment methodic made it possible to conduct the following experiments in IGR reactor: 1) Capsule testing without coolants of standard single fuel elements of power reactors in case of RIA accident, including testing with burned up fuel having energy release of 10-100 kJ/g ²³⁵ U. Medium in capsule is water, steam, air, helium, sodium, lead. 2) Single fuel elements capsule testing in conditions that simulate final accident phase with coolant loss. The veritable energy release from normal operating to residual was realized. As rule, steam was used as coolant or medium. 3)Fuel assembly testing of research water cooled reactors in simultaneous stopping of coolant circulation. IWW-2M reactor assembly experiments made possible verification of reactor operating limits in power variation conditions. 4) Standard fuel assembly testings of power reactors in conditions that simulated final LOCA accident phase with full switching reactor power sources off. The experiments are carried out on standard fuel assembly that consists of 18 WWER-1000 reactor type' of fuel elements. 5). The methodical stage with melt of reactor core materials mixture (fuel cladding, lattices of total mass of up to 1 kg) with subsequent melt drain on of substrate made of casing steel has bee carried out. 6) The experiments of melt of reactor core materials mixture with casing steel having subsequent melt drain into water has been carried out. 2 tabs

[en] The results of experiments dealing with measuring the radiation heating temperature under the effect of pulsed graphite moderated reactor fields realized for small-sized samples of tungsten, tantalum, molybdenum, chromium, copper, zirconium, nickel, iron, graphite, titanium, aluminium base alloy AMg-6, tin, lead, polyethylene and free thermocouples are discussed. The measuring system and approaches to thermocouples installation are described. The experimental results are considered. The influence of burst tail and delayed γ radiation on samples heating is demonstrated. Energy losses in the process of samples natural cooling are estimated. It is shown that heating temperature for some materials (tantalum, tungsten) during the reactor start-up with maximum power density level can achieve hundred degrees. The conclusion is made that this effect should be taken into account when preparing experiments and gained information analysis

[en] The results of experiments made in order to determine spectral composition of neutrons in the IGR impulse graphite moderated reactor experimental channel at fuel temperature close to the room one are considered. The set of activation and fission detectors with half-life period more than 0.5 days is applied for the neutron spectrum measurements. The algorithm based on the directed divergence method is used for reconstruction of neutron energy spectra in energy range of 0.6 eV - 18 MeV. The results of calculational studies into the influence of impurities in structural materials on portion of thermal neutrons in he spectrum in the channel centre are discussed as well. The conclusion is made that the calculational results agree well with experimental data

[en] The data of experiments realized in channels of the uranium-graphite pulsed reactor IGR for verification of the computer codes developed for simulating the so-called graphite transparency effects are analysed. The thermal and fast (E > 2.65 MeV) neutron fluences are measured in central and lateral reactor channels using activation detectors at reactor core initial temperature of the order of 300 K and fuel heating up to 1200 K. The transparency factor is calculated as the ratio of thermal neutron fluence in a pulse for heated reactor core to the fluence in a pulse at room temperature normalized for equal power releases in pulses. The conclusion is made that the measured results are in good agreement with the appropriate calculated values

No abstract available

[en] In the paper the STORKE automated system of licensing is considered. The product of group soft ware Lotus Notes was chosen in the capacity of platform for implementation of the STORKE system. The system is developing as information system is oriented on the operation with documents. The system is addressing for the Committee of Atomic Energy of the Republic of Kazakhstan and Institute of Atomic Energy of the National Nuclear Center of the Republic of Kazakhstan

[en] Problems concerned with the illegal turnover of nuclear materials and non-proliferation were considered on the example of identification of parts of the graphite-uranium reactor IGR. Identification of integral fuel parts does not require laborious and precision techniques. In the majority of cases we may restrict our consideration with inspection and measurement of geometrical dimensions. However, if reactor fuel part is destroyed or turned into scrap, to identify the 'unknown' fuel is it necessary to measure the uranium concentration in graphite, enrichment, U isotope composition, and composition of impurities, but this not always solve the problem. Therefore the choice and substantiation of the identification characteristics should be subjected to special investigations. (author)

[en] There were conducted in-pile experiments for grounding of realization possibility of the specified diagram of energy release rate in the experimental channels with model FA of fast reactor. There have been received data on reactivity effects caused by fuel replacement along reactor IGR central experimental channel. The data are necessary for grounding of safety of the experiments being planned. There have been performed researches on determination of interrelation between neutron flux distributing and fuel state in the central experimental channel. As fuel compactly bunched pellets of WWER-1000 type with enrichment of 4.4% by U²³⁵ were used. Measuring of neutron flux fluence was conducted with the use of activation detectors (static experiments without fuel location change) and small-size neutron chambers (dynamic experiments with fuel location change). There have been accomplished researches of impurity gases content in fuel of WWER-1000 type with enrichment of 4.4% by U²³⁵ . With this aim there have been conducted pressure measuring in the sealed cavity where fuel melting were being conducted. Energy release in fuel was provided at realization of controlled neutron impulse at IGR reactor

[en] It is discovered from calculation that the neutron lifetime is not a constant value, but depends on neutron generation point in the reactor core. An absorbent of control rods greatly effects on the life time value. The experimental outcomes confirming these features have been obtained. (author)