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Podol’sk, Russia

Pervov A.G.,Moscow State University of Civil Engineering | Andrianov A.P.,Moscow State University of Civil Engineering | Yurchevskiy E.B.,OKB Gidropress
Petroleum Chemistry | Year: 2015

It has been revealed that the use of reverse-osmosis (RO) membrane units for water treatment is complicated by the buildup of their concentrate waste needed to be disposed. A technology for recycling the concentrate, consisting of the removal (crystallization) of calcium carbonate present in the concentrate by seed crystals, has been proposed and experimentally tested. By this process using RO units, it is possible to obtain from natural water highly desalinated (softened) water and water with reduced concentrations of calcium and bicarbonate ions having the same salinity as the source water. The technology can be effectively used in the production of drinking water and integrated treatment of feed water for steam boilers and heat supply systems. © 2015 Pleiades Publishing, Ltd. Source


Steinbruck M.,Karlsruhe Institute of Technology | Birchley J.,Paul Scherrer Institute | Boldyrev A.V.,RAS Nuclear Safety Institute | Goryachev A.V.,RIAR | And 10 more authors.
Progress in Nuclear Energy | Year: 2010

This paper gives an overview on the status of knowledge of high-temperature oxidation of the two zirconium alloys Zircaloy-4 and E110 with special emphasis on results obtained during the SARNET period. The tin-bearing alloy Zircaloy-4 and the niobium-bearing alloy E110 are the materials for cladding and structures used in pressurised water reactors of the Western-type and VVERs and RBMKs, respectively. Results of separate-effects tests, single-rod tests, and large-scale bundle experiments are summarised. Focus is directed to oxidation kinetics at high temperature, hydrogen release and absorption by the remaining metal, and behaviour during quenching. Both materials behave very similarly as long as the superficial oxide scales formed during oxidation are dense and protective. Main differences are seen in connection with breakaway oxidation which leads to enhanced oxidation and hydrogen uptake and thus embrittlement and possibly earlier failure of the cladding. The temperature range at which pronounced breakaway is observed is different for the two alloys. The status of modelling of oxidation kinetics, thermo-mechanical behaviour during cooldown and the influence of irradiation are discussed at the end of the paper. © 2009 Elsevier Ltd. All rights reserved. Source


Ponomarenko G.L.,OKB Gidropress
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2013

Severe accident on the Fukushima NPP lowered the public trust for nuclear power engineering and must lead to the revision of the NPPs safety substantiation in the trend of the more strict regulatory body's requirements. This relates to the deterministic (DSA) and probabilistic (PSA) safety analysis. In the author's opinion, the most effective increase in the safety should be reached due to convergence or harmonization of deterministic and probabilistic methods. In particular, at fulfillment of DBAs in a number of cases one should consciously move in the direction of more realistic analysis of the actually BDBAs emergencies (in the traditional understanding) with the imposition of more than one Initiating Event (IE) or more than one Single Failure. In turn, when the PSA is fulfilled one should move from the realistic analysis without taking into account uncertainties in the direction of their reasonably conservative calculation. This work presents the practical realization of harmonious interaction of methods DSA and PSA based on the example of the estimation of the minimally sufficient CRs CPS for WWERs. A new conceptual approach to the definition of a minimum required quantity of CRs as part of the Spatial Effects Methodology is presented here and in papers [15, 16]. This approach combines the neutron-physical, thermal-hydraulic and probabilistic aspects, using DBC's Fuel Criteria generally accepted for CD realization. Analysis is performed with the up-to-date coupled code KORSAR/GP [1- 3] on example of the mode MSLB, which has the most dependence on EP with respect to any other accident. This paper presents an approach that allows: - to assess the EP success criteria with multiple stuck CRs for the purposes of PSA. In other words - assess the maximum number of stuck CRs when CD does not yet occur; - justify the low (P-EP fail ≤ 10-5, in accordance with Russian State Standard [4]) probability of EP failure on demand, for different amounts of CRs in a range from 49 till 121 pcs. Furthermore, this probability allows substantiate the low partial CD probability P-CDpartial < ∼10-10 for concrete scenarios with multiple stuck CRs as a negligible contribution in the total CD probability which should be P-CDtotal < 10 -5; - a joint optimization for positioning and quantities of CRs and power monitoring detectors (SPND) in central cell of FAs. Indeed, if the total number of CRs in WWER-1000 becomes more than 79-85 pcs, it creates some inconvenience, particularly the need to shift the power monitoring detectors from the central cell of fuel assembly into another noncentral cell. Copyright © 2013 by ASME. Source


Dzhalandinov A.,OKB Gidropress | Tsofin V.,OKB Gidropress | Kochkin V.,RAS Research Center Kurchatov Institute | Panferov P.,RAS Research Center Kurchatov Institute | And 5 more authors.
EPJ Web of Conferences | Year: 2016

Usually the synthesis of two-dimensional and one-dimensional discrete ordinate calculations is used to evaluate neutron fluence on VVER-1000 reactor pressure vessel (RPV) for prognosis of radiation embrittlement. But there are some cases when this approach is not applicable. For example the latest projects of VVER-1000 have upgraded surveillance program. Containers with surveillance specimens are located on the inner surface of RPV with fast neutron flux maximum. Therefore, the synthesis approach is not suitable enough for calculation of local disturbance of neutron field in RPV inner surface behind the surveillance specimens because of their complicated and heterogeneous structure. In some cases the VVER-1000 core loading consists of fuel assemblies with different fuel height and the applicability of synthesis approach is also ambiguous for these fuel cycles. Also, the synthesis approach is not enough correct for the neutron fluence estimation at the RPV area above core top. Because of these reasons only the 3D neutron transport codes seem to be satisfactory for calculation of neutron fluence on the VVER-1000 RPV. The direct 3D calculations are also recommended by modern regulations. © 2016 Owned by the authors, published by EDP Sciences. Source


Tikhomirov A.V.,OKB Gidropress | Ponomarenko G.L.,OKB Gidropress
International Conference on the Physics of Reactors 2012, PHYSOR 2012: Advances in Reactor Physics | Year: 2012

An additional verification of bundled software (BS) SAPFIR-95&RC [1] and code KORSAR/GP [2] was performed. Both software products were developed in A.P. Alexandrov NITI and certified by ROSTEKHNADZOR of RF for numeric simulation of stationary, transitional and emergency conditions of VVER reactors. A benchmark model for neutronics calculations was created within the limits of this work. The cold subcritical state of VVER - 1000 reactor stationary fuelling was simulated on the basis of FA with an increased height of the fuel column (TVS-2M) considering detailed presentation of radial and front neutron reflectors. A case of passing of pure condensate slug through the core in initially deep subcritical state during start of the first RCP set after refueling was considered as an examined condition of reactor operation. A relatively small size of the slug, its spatial position near the reflectors (lower and lateral), as well as failure of the inserted control rods of the control and protection system (CPS CR) to reach the lower limit of the fuel column stipulate for methodical complexity of a correct calculation of the neutron multiplication constant (Keff) using engineering codes. Code RC was used as a test program in the process of reactor calculated 3-D modeling. Code MCNP5 [3] was used as the precision program, which solves the equation of neutrons transfer by Monte-Carlo method and which was developed in the US (Los-Alamos). As a result of comparative calculations dependency of K eff on two parameters was evaluated - boron acid concentration (Cb) and CPS CR position. Reactivity effect was evaluated, which is implemented as a result of failure of all CPS control rods to reach the lower fuel limit calculated using the engineering codes mentioned above. Source

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