JSC Institute of Nuclear Materials

Sverdlovsk oblast, Russia

JSC Institute of Nuclear Materials

Sverdlovsk oblast, Russia

Time filter

Source Type

Kinev E.A.,JSC Institute of Nuclear Materials | Panchenko V.L.,JSC Institute of Nuclear Materials
Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika | Year: 2017

Radiation-induced swelling has a negative influence on the structural unit availability of the fast neutron reactor core. Therefore if to reduce swelling it is an important task to search for new steels and improve known ones. Since 2003 the 16Cr-15Ni-2Mo-Mn-Ti-V-B steel shows a significant increase in radiation resistance as a result of the improvement of composition and heat treatment. The swelling of 16Cr-15Ni-2Mo-Mn-Ti-V-B improved steel is studied with JSC INM's forces. The data about the maximum swelling temperature and the average speed of swelling in typical temperature ranges of the coolant and the dose rate of fast reactor was obtained. Research materials are based on the results of hydrostatic weighing and transmission microscopy measurements of steel samples density and swelling. Specific matters of hydrostatic measurement errors were discussed considering metallography data and immersion liquid choice. It was found that the average swelling rate of 16Cr-15Ni-2Mo-Mn-Ti-V-B improved steel under maximum swelling characteristic temperature is within the range of 0,04-0,14 %/dpa. There is a tendency of the characteristic temperature shift from 460 to 520°C as maximum damage dose increases from 60 to 80 dpa (1,3·10-6 and 1,7·10-6 dpa/s respectively). At low (less than 10 dpa) damage doses and minimum (less than 400°C) temperatures the swelling rate can reach 0,04 %/dpa. High-temperature metal corrosion causes hydrostatic measurement errors. According to the electron microscopy data, at temperature about 600°C and damage dose below 50 dpa, swelling rate does not exceed 0,01 %/dpa throughout the whole observation period.


Kozlov A.V.,JSC Institute of Nuclear Materials | Portnykh I.A.,JSC Institute of Nuclear Materials | Tselishchev A.V.,HIGH-TECH | Shilo O.B.,JSC Institute of Nuclear Materials | Asiptsov O.I.,JSC Institute of Nuclear Materials
Russian Metallurgy (Metally) | Year: 2014

Samples of 0.06C-16Cr-15Ni-2Mo-2Mn-Ti-Si-V-B and 0.07C-16Cr-19Ni-2Mo-2Mn-Ti-Si-V-P-B steels (ChS68 and EK164 steels, respectively) in the form of fuel element cladding tubes from a BN-600 reactor have been subjected to neutron irradiation in an IVV-2M research reactor to damage doses of 0.0015, 0.0050, and 0.0100 dpa at a temperature of 30°C. Based on a comparison of the results of dilatometric measurements of the irradiated samples and the samples in the initial state, the energies of vacancy migration in the steels are calculated. It is found that the energy of vacancy migration is 1.08 ± 0.02 eV in the ChS68 steel and 0.98 ± 0.02 eV in the EK164 steel. Using these values, the steady-state vacancy concentrations during irradiation of these steels in the BN-600 reactor are calculated. It is shown that the steady-state vacancy concentration in the EK164 fuel cladding portions irradiated in the lower half of the core is significantly lower than that in the ChS68 cladding. This is a cause of the higher resistance of the former steel to radiation-induced swelling as compared to that of the ChS68 steel upon irradiation in fast neutron breeders. © 2014 Pleiades Publishing, Ltd.


Tashlykov O.L.,Ural Federal University | Shcheklein S.E.,Ural Federal University | Luk'yanenko V.Y.,Ural Federal University | Mikhajlova A.F.,Ural Federal University | And 3 more authors.
Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika | Year: 2015

The aim of this work is to develop an algorithm of design of homogeneous composition of the radiation protective materials (RPM) for the optimization of radiation protection. Used for the investigation were the homogeneous radiation protective materials such as Abris, the production technology of which allows to obtain the desired concentration of fillers. The estimation of weakening ability of radiation protective material containing from 20 to 80% of barite, lead, tungsten, was carried out using high-precision calculation codes. To verify the results of calculations the experimental study of protective properties of Abris material with different concentrations of filling was carried out. Five sources of gamma radiation (60Co, 58Co, 198Au, 54Mn, 24Na) with characteristic energies of radiation were produced in IVV-2M research reactor for the experiment. A specially designed experimental installation and measuring device DKS-AT1123 were used. As a result of the research it was obtained the calculated dependency of radiation weakening coefficient for specific for different cases radioactive sources for the various compositions and thicknesses of the RPM. These data become initial for the optimization of radiation protection. Conclusions. 1. The design of RPM of homogeneous composition has considerable potential in the implementation of the radiation protection optimization principle. 2. A comparison oft he results of conducted studies of the gamma radiation dose weakening coefficient with homogeneous radiation protective materials of Abris RZ type depending on the composition and thickness showed that the difference between the experimental data and the values obtained by calculation does not exceed 5%. 3. The technology of production of Abris type homogeneous PRM allows to provide the required protective properties for the specific exposure conditions (composition of radioactive pollution).


Kinyov E.A.,JSC Institute of Nuclear Materials | Shikhalyov V.S.,JSC Institute of Nuclear Materials | Barybin A.V.,JSC Institute of Nuclear Materials
Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika | Year: 2015

The austenitic chromium-nickel steel EK164 is perspective material of the fast nuclear reactor rod cladding. The physical-chemical compatibility of this steel with typical pellet uranium dioxide nuclear fuel is important aspect for rod efficiency. The post reactor investigations of the combined rod assembling were carried out after peak burn-up 9,1 % FIFA and damage dose 77,3 dpa. There were rod claddings on basis of CHS68 and EK164 cold-worked steels to compose that assembling. The gamma-scanning, electrical-potential scanning and optical metallography methods was applied. The perilous regions of rod corrosion strengthening are high-temperature parts. This fact was previously established on basis of gamma-scanning and electrical-potential method data. The comparative analysis of the inner fuel rod corrosion for steel CHS68 and steel EK164 was carry out along core region height. The inner corrosion depth of steel CHS68 under contact with fuel has not exceeded 15 mm at the maximum power flux area by temperature below 540 °C. The same rod corrosion of steel EK164 has come to 10 mm at the core region centre. The maximum of corrosion damage both steels has detected at the range from 600 to 650 °C. It is amount less than 20 mm by means of both frontal and intercrystalline corrosion type. Essential differences of corrosion mechanism both steels haven't disclosed. The fact of local intension corrosion was detected near the fuel pellet joints and places of fission fragment caesium accumulation. Contrariwise the corrosion of EK164 steel was minimum within narrow gap between cladding and fuel where the caesium is absent. The peak cladding thinning of examined fuel rods have formed less than 5 % of initial thickness under fuel burn-up 9 % FIFA.


The phase composition and the characteristics of vacancy voids in cold-worked steel 07C–16Cr–19Ni–2Mo–2Mn–Ti–Si–V–P–B (CW EK164-ID) after neutron irradiation at damaging doses of 36–94 dpa and temperatures of 440–600°C are investigated. In the entire range of damaging doses and temperatures, voids with different sizes are observed in the material. The maximum void size increases with irradiation temperature up to ~550°C, whereas their concentration decreases. At higher irradiation temperatures, almost no coarse voids are observed. The concentration of fine voids (to 10 nm in size) sharply increases with temperature from 440 to 480°C. Further increases in the temperature do not result in the noticeable concentration growth. In the irradiation temperature range of 440–515°C, second phases precipitate (G phase, γ’ phase, and complex fcc carbides). At higher irradiation temperatures, there are Laves-phase particles, fine second carbides of the MC type, and needle shape precipitates identified as phosphides in the material. © 2016, Pleiades Publishing, Ltd.

Loading JSC Institute of Nuclear Materials collaborators
Loading JSC Institute of Nuclear Materials collaborators