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Kudinov P.,KTH Royal Institute of Technology | Davydov M.,Electrogorsk Research and Engineering Center
Nuclear Engineering and Design | Year: 2013

Ex-vessel termination of accident progression in Swedish type Boiling Water Reactors (BWRs) is contingent upon efficacy of melt fragmentation, quenching, solidification and formation of a coolable by natural circulation porous debris bed in a deep pool of water below reactor vessel. When liquid melt reaches the bottom of the pool it can cause formation of agglomerated debris and "cake" regions, which affect hydraulic resistance and thus coolability of the bed. This paper discusses development and validation of conservative-mechanistic and best estimate approaches to quantifying mass fractions of agglomerated debris at given conditions of melt release from the vessel. Fuel-coolant interaction (FCI) code VAPEX-P is used as a computational vehicle for modeling. Experimental data from the DEFOR-A (Debris Bed Formation and Agglomeration) tests with binary oxidic simulant material melt is used for validation of developed methods. The paper discusses the influence of different inherent uncertainties in the prediction of the fraction of agglomerated debris. © 2013 Elsevier B.V.


Kudinov P.,KTH Royal Institute of Technology | Davydov M.,Electrogorsk Research and Engineering Center
International Congress on Advances in Nuclear Power Plants, ICAPP 2014 | Year: 2014

Ex-vessel severe accident mitigation strategy in Nordic type Boiling Water Reactors (BWRs) imply that melt released into a deep pool of water below reactor vessel will form a coolable by natural circulation porous debris bed. However, if liquid melt is not completely fragmented and quenched when it reaches the bottom of the pool it can cause agglomeration of debris, increasing hydraulic resistance and thus worsening coolability of the bed. In the previous work we have developed and validated an approach to prediction of mass fractions of agglomerated debris using Fuel-Coolant Interaction (FCI) code VAPEX-P. This paper discusses development of a surrogate model (SM) which can predict fraction of agglomerated debris with high computational efficiency. Such model is a must for affordable sensitivity and uncertainty analysis in different accident scenarios. Details of the SM development and verification against full model are provided in the paper.


Melikhov V.,Electrogorsk Research and Engineering Center | Melikhov O.,Moscow Power Engineering Institute | Yakush S.,Russian Academy of Sciences | Rtishchev N.,Moscow Power Engineering Institute
Science and Technology of Nuclear Installations | Year: 2011

A specialized module VAPEX-M has been developed and implemented as a part of an integral code, SOCRAT, to enable the modeling of fuel-coolant interactions (FCIs) during severe accidents. The mathematical model and correlations for the main physical processes are described. Results of computational analysis of three experimental series reported in the literature are presented. The calculations were carried out by the combined SOCRAT/VAPEX code and were aimed at validation of the predictive capabilities of the code. The experiments chosen cover a wide range of physical parameters, which enables different aspects of the code to be verified, that is, drag correlations (MAGICO-2000), evaporation rate (QUEOS), fuel fragmentation, and interaction with the coolant in all complexity (FARO). Generally, reasonable agreement between the measured data and calculated results was obtained, which allows one to use the combined SOCRAT/VAPEX code for severe accidents analysis. Copyright © 2011 Vladimir Melikhov et al.


Melikhov V.,Electrogorsk Research and Engineering Center | Melikhov O.,Moscow Power Engineering Institute | Parfenov Y.,Moscow Power Engineering Institute | Nerovnov A.,Moscow Power Engineering Institute
Science and Technology of Nuclear Installations | Year: 2011

The horizontal steam generator (SG) is one of specific features of Russian-type pressurized water reactors (VVERs). The main advantages of horizontal steam generator are connected with low steam loads on evaporation surface, simple separation scheme and high circulation ratio. The complex three-dimensional steam-water flows in the steam generator vessel influence significantly the processes of the steam separation, distribution, and deposition of the soluble and nonsoluble impurities and determine the efficiency and reliability of the steam generator operation. The 3D code for simulation of the three-dimensional steam-water flows in the steam generator could be effective tool for design and optimization of the horizontal steam generator. The results of the code calculations are determined mainly by the set of the correlations describing interaction of the steam-water mixture with the inner constructions of the SG and interfacial friction. The results obtained by 3D code STEG with the usage of the different interfacial friction correlations are presented and discussed in the paper. These results are compared with the experimental ones obtained at the experimental test facility PGV-1500 constructed for investigation of the processes in the horizontal steam generator. Copyright © 2011 Vladimir Melikhov et al.


Blinkov V.N.,Electrogorsk Research and Engineering Center | Melikhov O.I.,Electrogorsk Research and Engineering Center | Melikhov V.I.,Electrogorsk Research and Engineering Center | Davydov M.V.,Electrogorsk Research and Engineering Center | And 2 more authors.
Science and Technology of Nuclear Installations | Year: 2012

In the frame of Tacis Project R2.01/99, which was running from 2003 to 2005, the bubble condenser system of Kola NPP (unit 3) was qualified at the integral test facility BC V-213. Three LB LOCA tests, two MSLB tests, and one SB LOCA test were performed. The appropriate test scenarios for BC V-213 test facility, modeling accidents in the Kola NPP unit 3, were determined with pretest calculations. Analysis of test results has shown that calculated initial conditions and test scenarios were properly reproduced in the tests. The detailed posttest analysis of the tests performed at BC V-213 test facility was aimed to validate the COCOSYS code for the calculation of thermohydraulic processes in the hermetic compartments and bubble condenser. After that the validated COCOSYS code was applied to NPP calculations for Kola NPP (unit 3). Results of Tacis R2.01/99 Project confirmed the bubble condenser functionality during large and small break LOCAs and MSLB accidents. Maximum loads were reached in the LB LOCA case. No condensation oscillations were observed. Copyright © 2012 Vladimir N. Blinkov et al.


Dombrovsky L.A.,RAS Joint Institute for High Temperatures | Davydov M.V.,Electrogorsk Research and Engineering Center
Computational Thermal Sciences | Year: 2010

This article is concerned with numerical modeling of thermal radiation from the zone of interaction of a melt jet with a water pool. This particular problem is a part of the analysis of complex interaction of the core melt with water in the case of a hypothetical severe accident in light-water nuclear reactors. The energetic contribution of thermal radiation has been studied in some detail in recent articles by the authors. In the present article, we focus on a solution related to possible optical diagnostics of the physical parameters of the process. These diagnostics can be based on comparison of the measured and calculated thermal radiation in the small-scale laboratory experiments. The sensitivity of the numerical data to some important parameters of the computational model is expected to be important to validate and improve the multiphase flow model. The radiation transfer model employed is based on the transport approximation. The numerical procedure includes ray-tracing calculations in the range of water semitransparency with a source function determined using the large-cell radiation model. It is shown that the visible radiation of the interaction zone contains important information on the process parameters, and these parameters might be identified on the basis of the developed computational procedure for the direct problem. © 2010 by Begell House, Inc.


Melikhov O.I.,Electrogorsk Research and Engineering Center | Elkin I.V.,RAS Research Center Kurchatov Institute | Melikhov V.I.,Electrogorsk Research and Engineering Center | Nikonov S.M.,Electrogorsk Research and Engineering Center | And 4 more authors.
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2014

In 2012 on the large-scale thermalphysic PSB-VVER facility modelling of three different type accidents is executed: - Guillotine rupture of the pipeline on a reactor inlet; - A small leak from "the cold" pipeline; - Rupture of pressurizer surge line. In experiments work of passive safety systems of new projects of NPP with RU VVER in the conditions of loss of all sources of an alternating current was modelled. The purpose of experiments was research of influence of new passive safety systems on a temperature condition of a fuel cladding. On the basis of the received experimental data verification of the Russian system codes TRAP-KS, KORSAR/GP and SOCRAT is executed. In paper the basic results of researches are presented. Copyright © 2014 by ASME.


Boltenko E.A.,Electrogorsk Research and Engineering Center
High Temperature | Year: 2016

The results of experimental studies of heat removal from convex and concave heating surfaces of annular channels with swirl and transit flow in the precrisis region are presented. The dependences describing the data on the heat removal rate from convex and concave heating surfaces are given. © 2016, Pleiades Publishing, Ltd.


Del Nevo A.,University of Pisa | Adorni M.,University of Pisa | D'Auria F.,University of Pisa | Melikhov O.I.,Electrogorsk Research and Engineering Center | And 5 more authors.
Science and Technology of Nuclear Installations | Year: 2012

The OECD/NEA PSB-VVER project provided unique and useful experimental data for code validation from PSB-VVER test facility. This facility represents the scaled-down layout of the Russian-designed pressurized water reactor, namely, VVER-1000. Five experiments were executed, dealing with loss of coolant scenarios (small, intermediate, and large break loss of coolant accidents), a primary-to-secondary leak, and a parametric study (natural circulation test) aimed at characterizing the VVER system at reduced mass inventory conditions. The comparative analysis, presented in the paper, regards the large break loss of coolant accident experiment. Four participants from three different institutions were involved in the benchmark and applied their own models and set up for four different thermal-hydraulic system codes. The benchmark demonstrated the performances of such codes in predicting phenomena relevant for safety on the basis of fixed criteria. Copyright © 2012 A. Del Nevo et al.


Bucalossi A.,EC JRC JRC F.5 | Del Nevo A.,ENEA | Moretti F.,University of Pisa | D'Auria F.,University of Pisa | And 2 more authors.
Nuclear Engineering and Design | Year: 2012

VVER 1000 reactors have some unique and specific features (e.g. large primary and secondary side fluid inventory, horizontal steam generators, core design) that require dedicated experimental and analytical analyses in order to assess the performance of safety systems and the effectiveness of possible accident management strategies. The European Commission funded project "TACIS 2.03/97", Part A, provided valuable experimental data from the large-scale (1:300) PSB-VVER test facility, investigating accident management procedures in VVER-1000 reactor. A test matrix was developed at University of Pisa (responsible of the project) with the objective of obtaining the experimental data not covered by the OECD VVER validation matrix and with main focus on accident management procedures. Scenarios related to total loss of feed water and station blackout are investigated by means of four experiments accounting for different countermeasures, based on secondary cooling strategies and primary feed and bleed procedures. The transients are analyzed thoroughly focusing on the identification of phenomena that will challenge the code models during the simulations. © 2012 Elsevier B.V.

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