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Dimitrovgrad, Russia

Varivtsev A.V.,JSC SSC RIAR | Zhemkov I.Y.,JSC SSC RIAR
Physics of Atomic Nuclei | Year: 2014

The application of the improved method for calculating the radiation heat generation in the elements of an experimental device located at the periphery of the BOR-60 reactor core results in a significant reduction in the discrepancies between the calculated and the experimental data. This allows us to conclude that the improved method has an advantage over the one used earlier. © 2014, Pleiades Publishing, Ltd. Source

Varivtsev A.V.,JSC SSC RIAR | Zhemkov I.Y.,JSC SSC RIAR
Physics of Atomic Nuclei | Year: 2014

The results of theoretical and experimental studies aimed at determining the radiation heat generation in the BOR-60 reactor reveal the drawbacks of the computational methods used at present. An algorithm that is free from these drawbacks and allows one to determine the radiation heat generation computationally is proposed. © 2014, Pleiades Publishing, Ltd. Source

Rogozkin B.D.,JSC VNIINM | Stepennova N.M.,JSC VNIINM | Fedorov Yu.Ye.,JSC VNIINM | Shishkov M.G.,JSC VNIINM | And 5 more authors.
Journal of Nuclear Materials | Year: 2013

In the article presented are the results of post-irradiation tests of helium bonded fuel pins with mixed mononitride fuel (U0.55Pu 0.45)N and (U0.4Pu0.6)N having 85% density irradiated in BOR-60 reactor. Achieved maximum burn-up was, respectively, equal to 9.4 and 12.1 at.% with max linear heat rates 41.9 and 54.5 kW/m. Maximum irradiation dose was 43 dpa. No damage of claddings made of ChS-68 steel (20% cold worked) was observed, and ductility margin existed. Maximum depth of cladding corrosion was within 15μm. Swelling rates of (U0.4Pu 0.6)N and (U0.55Pu0.45)N were, respectively, ~1.1% and ~0.68% per 1 at.%. Gas release rate did not exceed 19.3% and 19%. Pattern of porosity distribution in the fuel influenced fuel swelling and gas release rates. © 2013 Published by Elsevier B.V. Source

Kobylyansky G.P.,JSC SSC RIAR | Novoselov A.E.,JSC SSC RIAR | Obukhov A.V.,JSC SSC RIAR | Ostrovsky Z.E.,JSC SSC RIAR | And 3 more authors.
Journal of ASTM International | Year: 2011

Alloy E635 is used as a material for some parts of the water-water energy reactor (VVER)-1000 fuel assemblies (FAs). The evolution of structural components and redistribution of the elementary composition that occurred between the solid solution and phase precipitates in the E635 alloy were analyzed. The correlation between these changes and the irradiation-induced deformation of FA parts was determined. The common features of the E635 alloy's irradiation damageability was established as a result of the irradiation of model samples in the test reactor BOR-60 and the operation of products in the VVER-1000 reactor. The peculiarities of E365 alloy's irradiation damageability under high dose irradiation were revealed. The Laves phase (Zr(Nb,Fe) 2) particles were found to be the main type of secondary phase precipitates observed in the E635 alloy products with completely or partially recrystallized structures. Under irradiation, iron released from this phase causes a transformation of its crystalline structure, i.e., it changes from hexagonal-close-packed into body-centered-cubic with an Fe-enriched (-Nb phase. The Fe content in the particles decreases as the dose rises. The above transformations are not observed on those FA areas where the damage dose is low. Some bigger (up to 0.50μ) precipitates of the (Z,Nb) 2Fe phase with facecentered-cubic lattices were found in the material structure. Irradiation up to high damage doses results also in the appearance of secondary fine-dispersed (up to ∼5 in size) irradiation-induced precipitates. Practically no niobium is observed in the matrix, while all tin is in the solid solution and the Fe fraction in the matrix rises as the fluence becomes higher. The generation and formation of c-dislocations occur only near the β-Nb precipitates (former Laves phases); a-dislocations in the form of dislocation loops 10 nm in size and black dots are observed over the whole volume of recrystallized grains. On the whole, the changes in the structural-phase state of the FA parts tested in the VVER-1000 reactor correspond to the ideas about irradiation-induced damage in E635 alloy gained in experiments on irradiation of model samples in the BOR-60 reactor. Copyright © 2011 by ASTM International. Source

Burukin A.V.,JSC SSC RIAR | Izhutov A.L.,JSC SSC RIAR | Kobylyansky G.P.,JSC SSC RIAR | Kuznetsov V.I.,JSC VNIINM | And 5 more authors.
LWR Fuel Performance Meeting, Top Fuel 2013 | Year: 2013

The MIR reactor loop facilities (LFs) were used to test VVER fuel rods with different burnups under transient conditions. The purpose of the experiment was to get information about the behavior of fuel rods under irradiation and their state after tests and to create a database to improve and verify the calculation codes. Tests were done for both full-size fuel rods (FSFRs) removed from FAs spent at NPPs and refabricated fuel rods (RFRs). Some of RFRs were instrumented with gages for in-pile measurements. The paper presents data on the test conditions, changes in the fuel rod parameters recorded by in-pile gages (cladding diameter, elongation, fission gas release (FGR), fuel temperature) and PIE results. Source

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