No abstract available

[en] The survey of experiential investigations carried out on the IVG-1 reactor on program of carrying out of high-temperature and gas cooled space reactors in the period from1971 to 1988 and after the reconstruction in the period from 1990 to 1994 is given in the paper. IVG-1 reactor is research high-temperature gas cooled tank nuclear reactor of channel type on thermal neutrons with light water moderator and beryllium reflector of neutrons. Test fuel assemblies are placed within reactor core cells and its are cooled by gas coolant. It is possible the various fuel assemblies are tested in the reactor simultaneously. In the center of the reactor core the loop channel with diameter 164 mm is placed. During 1975-1988 in the reactor have been tested 4 fuel assemblies sets for nuclear rocket engines. A lot of technologic, constructive, and scientific and research works were carryied out on the rector IVG-1 by the Russian Scientific Centre Kurchatov Institute, the Physics and Energy Institute, the Scientific Production Corporation Luch and other institutions of Russia. The scientific and technical results were riched during test process, in particular the process possibility of fuel assemblies of nuclear rocket engines during 4000 secunds for medium specific power energy deposition in fuel elements equal to 20-25 kW/sm and hydrogen temperature up to 3100 K. The reactor IVG-1 is universal tool for carrying out of a lot of different investigations. 4 figs

[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] There was investigation of the state of rod-type carbide-graphite fuel elements with a high content of uranium at the mass level of 23-38%) after 70-hour testing in RADA high pressure (20 MPa) nitrogen ampoules of RA reactor. It was discovered that the radiation-chemical stability of the fuel elements in nitrogen greatly depend on the initial state of their structure and uranium content. Comparison of the quantitative factors of chemical and radiation-chemical stability of carbide-graphite fuel elements in nitrogen was made. (author)

[en] The basic objective of the work is to install the qualitative and quantitative indexes of the serviceability of the rod-type carbide fuel pins as applied to the exploitation conditions in the high temperature gas cooled reactor, where the nitrogen will be used as the coolant. For that purpose the state of the fuel pins, tested in the nitrogen technological channels of the ATS of IVG.1 (EWG-1) reactor in the series of four research nitrogen start-ups , including the energetic start-up with the duration of 160 s. and three standard start-ups with the duration of 510 s., was researched. On the base of the results of the post-reactor research of the fuel pins of ten channels of ATS it is determined that the ceramic fuel pins of (U, Zr)C+C, (U, Nb)C+C and (U, Zr,Nb)C are enough serviceable in the severe conditions of the high temperature tests in the flow path of the chemically aggressive coolant. The lack of the surface cracks in the fuel pins, lack of the fuel pins failures and lack of overweight and thickening of the fuel pins are revealed. It is observed the oxy-nitration of the fuel pins surfaces (at appearance of the characteristic color tones and presence of the slight burning of the fuel rods to each other), however, the depth of the oxy-nitration , even of the fuel pins of the output heating sections, tested at 2800 K, did not exceed 10 μm. It is found out that the levels of the radioactive change of the fuel pins parameters are the same as of the fuel pins of the hydrogen technological channels, tested at the same temperatures and up to the same neutrons fluences. The low change of the fuel pins strength is observed; the strengthening of the fuel pins on the output heating sections for ∼ 20 % (due to the appearance of the residual radiation macro stresses) and weakening of the fuel pins in the output sections for ∼30 % (due to oxy-nitration and erosion of the surface, and also non-congruent evaporating of the surface material). The prognostic analysis of the fuel pins behavior in the case of the increase of the number of the reactor start-ups from 4 to the reserve 40 and duration of the tests from ∼700 s to reverse 5000 s is made. It is shown that their strength can reduce under the influence of the oxy-nitration, erosion, evaporating, corrosion under the pressure and tiredness of the vibrational load down to hazardous low level. That's why the first principal decision of the problem of increasing the serviceability of the ceramic fuel pins in the nitrogen - cooled channels is to change the carbide fuel pins to the carbonitride ones. The second decision of the problem is to remove the curling of the bundles of the fuel pins in the heating sections. In that case the favourable situation is created, in which the maximal depth of the fuel pins oxy-nitration will be of the principal value, and this value, according to the design estimations, is not to exceed 30 μm by the end of the full-resource tests. Taking into account this fact, the conclusion of the possibility to use the rod carbide fuel pins in the nitrogen cooled reactor without changing their material composition is made

[en] In the IGR reactor have been made of experimental testing of interaction composition materials of reactor core with water for WWER type reactor. Composition of materials presents mass 1 kg, including short staff WWER reactor's fuel elements with stainless steel smelting in chamber. Melting of composition is provided by working the reactor under impulse regime. The data about energetic parameters of interaction of corium with water (pressure impulse, degree of dispersive melt, conversion ratio), hydrogen yield contain and structure of melt fragment and data about yield dynamic and composition of fission products have ben received in consequence of the investigation

[en] Aim of this work is definition of the carbide fuel pins radiation-chemical stability with uranium content at the 15 % (mass) level in the low pressure nitrogen (about 1 MPa) in the condition of irradiation at ∼1000 K in the flow technological channels of the EWG.1 reactor. It is determined, that after 9.5 hour irradiation in the nitrogen flow all fuel pins keep an integrity and an essential portion of the initial strength

[en] The principle results of carbide fuel rods testing during series of IVG.1 reactor starts up at regime simulating nuclear engine regime of nuclear moving power unit are given. Considerable degradation of initial fuel elements status increasing from start up to start up and which could resulted fail of separate technological channels is shown. Origin case of extreme degradation of fuel elements status are insufficient thermal strength of fuel elements operation in the field brittle state of sintered carbide material, Possible ways of artificial reinforce of fuel elements of low temperature sections, increasing its thermal strength up to required level

[en] Radiation stability of fuel elements from (U, Zr)C+C and (U, Zr, Nb)C at irradiation in low flow ampoule reactor RA is studied. The examination was carried out under fuel elements operation conditions simulating the low power energy regime for reactor of two-regime nuclear moving power unit. It is established high rate of function capacity for work of the carbide fuel elements at given conditions of reactor testing with total duration more than 4,500 hours

[en] Radiation stability of ceramic fuel elements made from (U,Zr)C+C and (U+Zr,Nb)C containing U-235 from ∼9 to ∼18% (mass) under irradiation in the capsules of RA low flux reactor is studied. During the experiment the condition simulating operation condition of these fuel elements under power generating mode at low power of two-mode nuclear power engine (NPE). Fuel elements made by Podolsk State Scientific Research Institute Scientific Production Association 'Luch' in shape of wounded by length bars 100 mm long with the width of a blade equal to 1,24 mm and diameter 2,2 mm were tested in highly assembled bunches in heating sections (HS); each capsule contained 7 HS with 151 fuel element in each. Testing condition were: radiation intensity ∼1,5·¹² thermal neutrons/(sm²·s), fuel elements temperature - in the range of 900-1700 K, coolant -highly pure helium under the pressure of ∼0,45 MPa. In the course of testing 14 capsules with duration on irradiation from 1 to 4525 hours were dismantled. Results of post reactor studies of the condition of fuel element from these capsules demonstrated high radiation resistance of carbide cladding-free fuel rods under the low power condition of NRE. Low swell (<1,5 % at T_irrd=1200 K) and insignificant rod deformation (bending) that was revealed when irradiation period was higher than ∼3000 hours. 20-40 % strengthening of fuel elements for irradiation period more than 10 hours was found. It was demonstrated that radiation strengthening does not depend on irradiation temperature and does not disappear after long time annealing at T>T_irrd and is a result of healing of technological defects in sintered fuel carbide simulated by irradiation