Dimitrovgrad, Russia
Dimitrovgrad, Russia

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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.


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.


Izhutov A.L.,JSC SSC RIAR | Krasheninnikov Y.M.,JSC SSC RIAR | Zhemkov I.Y.,JSC SSC RIAR | Varivtsev A.V.,JSC SSC RIAR | And 3 more authors.
Nuclear Engineering and Technology | Year: 2015

The fast neutron reactor BOR-60 is one of the key experimental facilities worldwide to perform large-scale tests of fuel, absorbing, and structural materials for advanced reactors. The BOR-60 reactor was put into operation in December 1969, and by the end of 2014 it had been operating on power for ~265,000 hours. BOR-60 still demonstrates potential capabilities to extend the lifetime of sodium-cooled fast reactors. The BOR-60 lifetime should have expired at the end of 2014. Over the past few years, a great scope of work has been performed to justify the possibility of extending its lifetime. The work included inspection of the equipment conditions, calculations and experimental research on operating parameters and the conditions of nonremovable components, investigation of the structural material samples after their long-term operation under irradiation, etc. Based on the results of the work performed, the residual lifetime was evaluated and the reactor operator made a decision to extend the lifetime period of the BOR-60 reactor. After considering both a set of documents about the reactor conditions and the positive decision of independent experts, the Regulatory Authority of the Russian Federation extended the BOR-60 operating license up to 2020. © 2015, Published by Elsevier Korea LLC on behalf of Korean Nuclear Society.


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.


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.


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.


Osipova T.A.,Dimitrovgrad Engineering Technological Institute NRNU MEPhI | Valishin M.F.,JSC SSC RIAR | Uzikov V.A.,JSC SSC RIAR | Palachyov P.S.,JSC SSC RIAR
Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika | Year: 2015

The current research work provides the results of the calculation analysis of the thin% walled samples cooling during reactor irradiation in a double%hulled ampoule channel with a natural convection supply of the heatsink. Making the ampoule channel in the form of a double%hulled construction enables changing thermal resistance of the channel wall thus regulating heat leak out of it by gas composition and pressure in the cavity between the hulls. The aim of the research is to identify possible regimes of sample cooling in the present channel. The calculation analysis was carried out using the thermal%hydraulic code RELAP5/MOD3.2. In the calculations helium and nitrogen are seen as a gas gap between the hulls. There is a demonstration of the main regularities of changing irradiation temperature regimes in relation to the capacity of energy release in the channel construction materials and the irradiation device, the height of the circulation loop and thermal resistance of the channel hull. Varying the height of the circulation loop and the capacity of energy release in the construction materials enables cooling sample regimes in the range from the temperature of the first contour coolant to boiling temperature at a given pressure (50 - 331°C). Without boiling of coolant on the samples at the maximum (8 m) height of the circulation loop at most 55 kW (14 W/g on the samples) is allocated using helium as a channel cavity gas, exploiting nitrogen - 15 kW (3,7 W/g on the samples); at a minimum (1 m) height of the circulation loop no more than 10 kW (2,5 W/g on the samples) and 5 kW (1,2 W/g on the samples), correspondingly.


Markov D.V.,JSC SSC RIAR | Pavlov S.V.,JSC SSC RIAR | Novoselov A.Ye.,JSC SSC RIAR | Polenok V.S.,JSC SSC RIAR | And 4 more authors.
LWR Fuel Performance Meeting/Top Fuel/WRFPM 2010 | Year: 2010

The post-irradiation examinations results of the new-generation VVER and RBMK fuel are presented. Assessment of spent fuel assemblies state have been conducted concerning the main parameters effecting their operational reliability: fuel assembly geometrical stability in general, change in geometry and corrosion state of fuel rods and skeleton components, change in material microstructure of skeleton components, cladding and fuel.


Burukin A.V.,JSC SSC RIAR | Izhutov A.L.,JSC SSC RIAR | Ovchinnikov V.A.,JSC SSC RIAR | Markov D.V.,JSC SSC RIAR | And 4 more authors.
LWR Fuel Performance Meeting/Top Fuel/WRFPM 2010 | Year: 2010

This paper provides data on testing of the high-burnup VVER-440 fuel rods under power cycling conditions in the MIR reactor. The purpose of testing was to obtain experimental data on their serviceability (particularly, data on fission gas release (FGR), fuel temperature and degree of pellet-cladding mechanical interaction (PCMI) as well as characteristics of fuel rods after their irradiation. Results obtained during testing of the VVER-440 fuel rods with a burnup higher than ∼50 MWd/kgU under power cycling conditions over the range 100-60-100% as well as PIE results verified their serviceability and high reliability. Information about conditions and results of the tests as well as PIE results is used to estimate serviceability of the high-burnup VVER-440 fuel rods under similar operating conditions and verify calculation codes.


Varivtcev A.V.,JSC SSC RIAR | Zhemkov I.Yu.,JSC SSC RIAR | Boev A.V.,JSC SSC RIAR | Ishunina O.V.,JSC SSC RIAR | And 3 more authors.
Izvestiya Wysshikh Uchebnykh Zawedeniy, Yadernaya Energetika | Year: 2016

Tests of perspective materials and new generation reactor core elements are conducted in the BOR-60 reactor. An irradiation rig (IR) with a fuel heater is proposed to be used for reactor tests of different structural materials at the high temperatures. This type of IR has a number of advantages in comparison with IR containing thermo-insulated capsules that is often used now. Calculations and experimental investigations of the IR were performed in the core of the BOR-60 reactor. Results of the special methodical experiment confirmed the possibility of providing the necessary temperature conditions for the samples during irradiation. The neutronics were calculated by MCU-RR code and the thermal hydraulic characteristics were calculated using ANSYS CFX complex. Comparison of calculated temperature values with experimental ones showed that they correlate well - the difference between them is less than experimental error. It means that the used computer codes, models and methods are suitable for such calculations. The calculations and experiments were carried out to see the temperature distribution in the IR with a fuel heater irradiated in the BOR-60 and then put into the dry storage channel. Decay heat rate values inside the IR fuel pins calculated by AFPA code, and the temperatures were calculated using ANSYS CFX. The performed calculations and experiments also showed it possible to remove the IR with a fuel heater from the reactor on the second day of outage, the limited cladding temperatures and required samples temperatures being not exceeded.

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