[en] Results of investigations into the sensitivity of the KtV-type small-sized triaxial ionization chambers and cable type Compton emission detectors to gamma radiation are presented. The detectors under consideration are included into automated systems for power distribution, monitoring and control in nuclear reacor cores. Investigations are performed in the RK-LM in-pile loop operating in the IRT-M reactor. Gamma radiation dose rate corresponds to 10⁵ g.eq. of radium. The liquid metal alloy containing 25% indium+62% gallium+13% tin is used as a gamma source. The detector sensitivity calibration is performed without activation of their structural materials with mean-square error being no more than ±4%. The obtained data show that ionization chamber sensitivity to gamma radiation under reactor and loop conditions agree well enough with each other. Ionization chamber sensitivity difference from one another makes up no more than 30%. Emission detector sensitivity spread from the mean value does not exceed ±10%

[en] In order to carry out experimental studies in the field of nuclear physics and reactor technology involving a high flux density of neutrons (up to 10E18/sec/cm²), a heat capacity type impulse graphite reactor (IGR) with an integral neutron fluence of 1.1E17/sec/cm² (after complete withdrawal of the rods) is used. The reactor works in two main regimes, viz., self-quenching bursts with introduction of an excess reactivity up to some β_ef at a minimum half-width pulse of 0.1 sec and a controlled regime with a pulse width extending over several seconds. The open-quotes criticalclose quotes regime (physical start-up of the reactor) precedes each of these regimes. The experience gained during the operation of IGR shows that when the control and protection system receives signals from the regular lateral ionization chambers that are located at a distance of more than 1 m from the central channel, one observes differences between the desired regularity of variation of the power of the reactor and the realized regularity of variation of the neutron flux (energy release) in it. The ratio of the chamber current and the neutron flux density in this chamber can vary up to 50% because of the difference in the temperature of the neutrons in the region of location of the chambers and the channel itself during the heating process of the graphite lining and due to the effect of control rod displacement

[en] This report describes work carried out within the cooperation programme between the Norwegian Radiation Protection Authority and the Directorate of State Supervision for Nuclear and Radiation Safety of the Ministry of Defense of the Russian Federation performed in 2008-2009. It focuses on development of improved regulatory documents and supervision procedures for handling spent nuclear fuel and radioactive waste at facilities that are no longer used by the Russian Federation Navy but that are still under military supervision and control. (Author)

[en] Six compton emission detectors of cable type with emitters containing hafnium and gadolinium and four triaxial small-sized fission chambers are tested in the experimental channels of the IGR pulsed graphite moderated reactor. Undisturbed thermal neutron flux density varied within the range of 5x10¹³-5x10¹⁵ cm^-2xc^-1, fast neutron flux density made up 1.7x10¹²-1.7x10¹⁴ cm^-2xc^-1, γ radiation dose rate were 36-360 C/(kGxs). Tests were conducted periodically during a year. The thermal neutron fluence accumulated during the tests did not exceed 10¹⁸ cm^-2. The data obtained confirms practically absolute inertia-free character of investigated detectors which provides for their high-efficient application in automated power distribution control systems for nuclear power reactors

[en] A survey is made of a former shore-based technological base of the naval fleet, transferred to the Ministry of Atomic Energy for ecological rehabilitation. The main sources of nuclear and radiation danger are determined. It is shown that a spontaneous chain reaction is possible when degraded fuel is removed from water-filled cells. A technology is proposed for safe removal. Combining the confrontational and experimental studies made possible to perform quickly and with small dose expenditures work in radiation and dangerous locations which are difficult to access and make recommendations for decreasing the γ-ray dose rate

[en] A design of the test fuel assembly of MR type for in-pile tests to determine design safe operation margins is presented in the paper. The tests will be performed in the pulsed uranium-graphite reactor of IGR type in Semipalatinsk (Kazakh Republic) under emergency models of a loss-of-coolant accident and a reactivity initiated accident. (author)

[en] In the course of testing of fuel elements (FE) and fuel assembly (FA) behaviour under transient and emergency conditions during capsule and loop test at impulse uranium-graphite reactor IGR, it is essential to provide reliable monitoring of hydrodynamic, thermophysical and neutronic parameters. These test are particular for high-speed processes and the wide range of controlled parameters. The topics covered by the work are as follows: -development of the technique for neutron calculations; -techniques and devices to measure parameters of items under testing; -calculation and experimental justification and prediction of the physical parameters and test modes. Routine neutronic monitoring (determination of fuel elements (FA) power, power profile and amplitude, integral power density over start-up) is based on the use of signals from in-core detectors and result of the neutronic calculation. The equation to estimate FA power on the base of signals from in-core detector is given. Small-size high-temperature ionisation chambers of KtV and Compton emission neutron detectors were used as in-core detectors. In-core detectors are placed inside the central experimental channel of the reactor in the immediate vicinity to the object under testing

[en] The loop reactor facility UPSh constructed in 1988 on the base of the pulse uranium-graphite reactor IGR-1 is provided with necessary equipment and system allowing to perform investigation within a large-scale range of parameters. The facility is used to carry out the works enabling to perform comparative studies of various fuel elements (FE) and fuel assemblies (FA) serviceability. There are determined permissible variations in power change rate with registered FE conditions at different loading rates. As particular tasks has bean studied dryout occurring in both subcooled water and boiling system, the dynamic modes related with flow rate turnover, their progression at different FA power levels and getting natural circulation. The loop reactor facility based on IGR-1 reactor allows to provide adequacy to natural conditions, first in terms of speed and amplitude of changed parameters. Some of variable parameters such as flow rate , pressure, coolant temperature at FA inlet are implement by way of layout engineering of the loop facility and equipment while other parameters like speed and the value of FA power charge - by using corresponding variation in IGR-1 reactor power. The basic parameters of IGR-1 reactor and UPSh loop facility are presented. 1 tab

[en] Procedure of the research on behaviour of melt was tested by means of simple model experiments using paraffin with boric acid addition. Calculation of neutron-physical characteristics of assemblies was conducted by the PRIZMA D and MCNP programs. Analysis of the obtained experimental results allows number of detectors and their disposition to be optimized. Possibility for the generation of information on transfer of molten mass by means of small thermal neutron detectors is demonstrated

[en] Current of sensible units of 5-sectional compton emission neutron detector located at different levels through the height of the central experimental channel of pulse graphite reactor (PGR) during controlled pulse heating reactor fuel up to 1400 K was measured. It is established that the configuration of thermal neutrons distribution through the height of the channel in the course of start changes - at room temperature of fuel a maximum of the distribution locates at close range to the reactor core center. As increasing temperature of fuel up to 550 K and removal of controlled rods it is shifted down around 13 cm, during further heat-up to ∼ 1400 K and removal of rods it reverts to the level of center