OKB Gidropress

Podol’sk, Russia

OKB Gidropress

Podol’sk, Russia
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Deev V.I.,National Research Nuclear University MEPhI | Kharitonov V.S.,National Research Nuclear University MEPhI | Churkin A.N.,OKB GIDROPRESS
Thermal Engineering | Year: 2017

Experimental data on heat transfer to supercritical pressure water presented at ISSCWR-5, 6, and 7 international symposiums—which took place in 2011–2015 in Canada, China, and Finland—and data printed in recent periodical scientific publications were analyzed. Results of experiments with annular channels and three- and four-rod bundles of heating elements positioned in square or triangular grids were examined. Methodology used for round pipes was applied at generalization of experimental data and establishing of correlations suitable for engineering analysis of heat exchange coefficient in conditions of strongly changing water properties in the near-critical pressure region. Empiric formulas describing normal heat transfer to supercritical pressure water mowing in annular channels and rod bundles were obtained. As compared to existing recommendations, suggested correlations are distinguished by specified dependency of heat exchange coefficient on density of heat flux and mass flow velocity of water near pseudo-critical temperature. Differences between computed values of heat exchange coefficient and experimental data usually do not exceed ±25%. Detailed statistical analysis of deviations between computed and experimental results at different states of supercritical pressure water flow was carried out. Peculiarities of deteriorated heat exchange were considered and their existence boundaries were defined. Experimental results obtained for these regimes were generalized using criteria by J.D. Jackson that take the influence of thermal acceleration and Archimedes forces on heat exchange processes into account. Satisfactory agreement between experimental data on heat exchange at flowing of water in annular channels and rod bundles and data for round pipes was shown. © 2017, Pleiades Publishing, Inc.

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.

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.

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.

Nikolaeva A.,OKB GIDROPRESS | Skibin A.,OKB GIDROPRESS | Krutikov A.,OKB GIDROPRESS | Golibrodo L.,OKB GIDROPRESS | And 3 more authors.
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2015

Most accidents involving hydrogen begin with its leakage and spreading in the air and spontaneous detonation, which is accompanied by fire or deflagration of hydrogen mixture with heat and/or shocks, which may cause harm to life and equipment. Outflow of hydrogen in a confined volume and its propagation in the volume is the worst option because of the impact of the insularity on the process of detonation. According to the safety requirements for handling hydrogen specialized systems (ventilation, sprinklers, burners etc.) are required for maintaining the hydrogen concentration less than the critical value, to eliminate the possibility of detonation and flame propagation. In this study, a simulation of helium propagation in a confined space with different methods of injection and ventilation of helium is presented, which is used as a safe replacement of hydrogen in experimental studies. Five experiments were simulated in the range from laminar to developed turbulent with different Froude numbers, which determine the regime of the helium outflow in the air. The processes of stratification and erosion of helium stratified layer were investigated. The study includes some results of OECD/NEA-PSI PANDA benchmark and some results of Gamelan project. An analysis of applicability of various turbulence models, which are used to close the system of equations of momentum transport, implemented in the commercial codes STAR CD, STAR CCM +, ANSYS CFX, was conducted for different mesh types (polyhedral and hexahedral). A comparison of computational studies results with experimental data showed a good agreement. In particular, for transition and turbulent regimes the error of the numerical results lies in the range from 5 to 15% for all turbulence models considered. This indicates applicability of the methods considered for some hydrogen safety problems. However, it should be noted that more validation research should be made to use CFD in Hydrogen safety applications with a wide range of physical effects involved. Copyright © 2015 by JSME.

Thermal Engineering (English translation of Teploenergetika) | Year: 2014

A transition from the mass balance equations based on Kirchhoff’s first and second laws to modeling on the basis of a discretized continuity equation is made for describing a hydraulic network. A technique for calculating high-dimension hydraulic and heat networks based on the numerical finite-difference control volume method is developed. Unlike the existing approaches, the proposed technique does not involve the need to determine hydraulic loops and boils down to solving the problem of obtaining a unified field of pressures for the entire calculation region. This advantage of the proposed method opens the possibility of applying it for solving high-dimension problems containing more than a million of hydraulic links. The proposed numerical method features stable operation for hydraulic networks the neighboring links of which may have pressure drop coefficients differing from each other by more than 10 orders of magnitude. In contrast, the global gradient algorithm implemented in the standard software system EPANET is of little use for such applications. The convergence rate of the proposed technique is close to that of the Todini gradient algorithm and is almost independent of the problem dimension. © 2014, Pleiades Publishing, Inc.

Makarov V.V.,OKB GIDROPRESS | Afanasiev A.V.,OKB GIDROPRESS | Matvienko I.V.,OKB GIDROPRESS | Puzanov D.N.,OKB GIDROPRESS | And 2 more authors.
LWR Fuel Performance Meeting, Top Fuel 2013 | Year: 2013

The number of the measures to implement the principle of zero nuclear fuel failure includes finding a solution to the problem of enhancing the operational reliability of fuel assemblies (FA). A failure is determined as a loss of fuel integrity or its mechanical damage that impede its further operation. In its background the task of enhancing one strength aspect or another (thermomechnical, vibrational, seismic) was initiated in the effort to improve the economic characteristics, highlighted by failures and solved as soon as the problems arose. Since certain knowledge is still not available at such an approach a conservative increase in the resistance to one factor can lead to strength decrease for another factor by an unknown value. For the VVER-1000 FAs the task of increasing the thermo-mechanical strength arose after the thermo-mechanical bowing of FA and rod cluster control assembly (RCCA) sticking, the problem of providing the fretting-wear resistance appeared when steel spacer grids (SG) were replaced with zirconium ones and the SG-to-SG spans were made longer and the task of FA seismic strength came into view after the Fukushima NPP accident. The implementation of a zero failure principle implies fuel unloading and termination of leaky fuel operation. In a more general case there is a task to ensure the FA integrity for acceptable budget and time. FA integrity can be interpreted as the retention of the parameters of the FA geometrical shape (design-basis limits not exceeded) and also the fuel rod tightness sufficient to provide the nuclear and radiation safety in the course of FA life cycle. The cycle comprises the manufacturing, operation under normal conditions and at anticipated operational occurrences, fuel handling procedures and recycling. The paper contains the analysis of the results of a few cycles of experimental studies performed to justify the FA integrity at the stage of designing. The example of solving some problems in the area of zirconium material tribology and FA dynamic deformation at seismic impact shows the possibility to reduce conservatism in some design solutions and improve the operational and economic indices of fuel. Further work in this direction can help elaborating a Code of Rules and Recommendations for FA design development. The database on the tests for all the kinds of impacts, is definitely supposed to become part of the Code of Rules and allow a designer to optimize the necessary FA quality.

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.

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.

International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2013

Paper [1] describes a unique method and basic results of coolant mixing experimental study, carried out on the working power unit WWER-1000 of "Bushehr" NPP. Method is characterized by the use of an emergency boron insertion system for creating the nonuniform distribution of indicator (tracer), and also by the use of all regular systems of neutron and temperature monitoring for increasing the authenticity. Method showed high reliability, effectiveness, availability and flexibility. The corresponding application for the invention of a new method is applied for a patent. Present work supports the conclusions of the paper [1] and contains the more detailed additional information and results according to the mixing indices, and also to the dissipation of temperature indicator in different sections of circulation loop. Some results of post-test modeling by calculation code KORSAR/GP are presented too. Spatial and temporal detailing relate also to the dynamics of appearance, stabilization, propagation and disappearance of boron and boron-free "spots" in the core with start-up and trip of the emergency boron insertion pumps. Copyright © 2013 by ASME.

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