Institute of Reactor Materials

Sverdlovsk oblast, Russia

Institute of Reactor Materials

Sverdlovsk oblast, Russia
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Kozlov A.V.,Institute of Reactor Materials | Portnykh I.A.,Institute of Reactor Materials | Blokhin A.I.,RAS Institute of Physics and Power Engineering | Blokhin D.A.,RAS Institute of Physics and Power Engineering | Demin N.A.,RAS Institute of Physics and Power Engineering
Inorganic Materials: Applied Research | Year: 2013

The mechanism of void nucleation in ChS-68 austenitic steel induced by irradiation in a fast neu-tron reactor associated with formation of the complexes of helium vacancies and transmutation helium atoms is considered. The temperature dependence of the critical size of voids is calculated for the neutron radiation spectrum of the BN-600 fast breeder reactor. The data obtained are compared with the results of transmission electron microscopy studies of ChS-68 austenitic steel irradiated as a fuel cladding in the BN-600 reactor in the temperature range of 380-580°C. © Pleiades Publishing, Ltd., 2013.

Kinev E.A.,Institute of Reactor Materials | Tsigvintsev V.A.,Institute of Reactor Materials
Atomic Energy | Year: 2013

The article examines the interaction of fuel with ferrite-martensite steel fuel-element cladding under fast reactor conditions. EP-450 steel cladding curbs the deformation of the kernel in irradiated fuel elements at the end of an operating period. The linear thermal expansion coefficient of ferrite-martensite steel is 30% lower than that of austenitic steel and is comparable to that of sintered uranium dioxide. EP-450 steel is not subject to radiation swelling in the damaging dose range studied. Thus, the technological fuel-cladding gap vanishes at burnup above 4-5% h.a. The fuel can exert a high mechanical pressure on the cladding and cladding can become deformed as a result of creep. Under the conditions of cladding-limited swelling uranium dioxide undergoes creep. Plastic flow of the central regions of the fuel pellets in fuel assemblies in the high-enrichment zone causes considerable contraction of the central opening. The competing process of axial mass transfer, which enlarges the kernel cavity at the center of the active part, does not develop consistently.

Russkikh I.M.,Institute of Reactor Materials | Seleznev E.N.,Institute of Reactor Materials | Tashlykov O.L.,Ural Federal University | Shcheklein S.E.,Ural Federal University
Physics of Atomic Nuclei | Year: 2015

The significance of optimizing the content of components of a radiation-protective material, which is determined by the isotopic composition of radioactive contamination, depending on the reactor type, operating time, and other factors is demonstrated. The results of computational and experimental investigation of the gamma-radiation attenuation capacity of homogenous radiation-protective materials with different fillers are reported. © 2015, Pleiades Publishing, Ltd.

Rodchenkov B.S.,Moscow Power Engineering Institute | Evseev M.V.,Institute of Reactor Materials | Strebkov Y.S.,Moscow Power Engineering Institute | Sinelnikov L.P.,Institute of Reactor Materials | Shushlebin V.V.,Institute of Reactor Materials
Journal of Nuclear Materials | Year: 2011

The high strength (α + β) Ti-6Al-4V alloy was selected as the material for flexible attachments of the shield blanket modules in the ITER reactor. The different technologies used for manufacturing this alloy are: forging, stamping or pressing. The microstructures resulting from these processing methods can vary significantly and as a consequence the properties, including irradiation behavior, also vary. There are limited data available on the irradiation behavior of these materials. Specimens cut in the longitudinal and transversal directions of forged and stamped material were studied, with some of the specimens hydrogen charged to ∼400 ppm H2. In the unirradiated condition the forged alloy had slightly more ductility than the stamped alloy. The strength properties of both were practically the same. Neutron irradiation of these materials in the IVV-2M reactor reached doses of ∼0.2 and 0.3 dpa at temperatures 240-260 °C. Irradiation resulted in substantial hardening and significant decrease of the fracture toughness of specimens from both materials. © 2010 Elsevier B.V. All rights reserved.

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