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  • 1.
    Almjashev, V.I.
    et al.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Barrachin, M.
    Institut de Radioprotection et Suˆrete´ Nucle´aire (IRSN).
    Bechta, Sevostian
    AP Aleksandrov Res Inst Technol, Sosnovyi Bor 188540, Russia.
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Defoort, F.
    Laboratoire de Physico-chimie et Thermohydraulique Multiphasiques (LPTM), CEA/Grenoble, DTN/SE2T/LPTM – 17 rue des Martyrs.
    Fischer, M.
    Framatome ANP GmbH.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Hellmann, S.
    Framatome ANP GmbH.
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Lopukh, D.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Mezentseva, L.P.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Miassoedov, A.
    Institute for Nuclear and Energy Technologies, Forschungszentrum Karlsruhe.
    Petrov, Yu.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Vitol, S.A
    A.P. Aleksandrov Research Institute of Technology.
    Eutectic crystallization in the FeO(1.5)-UO(2+x)-ZrO(2) system2009In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 389, no 1, p. 52-56Article in journal (Refereed)
    Abstract [en]

    Results of the investigation of the FeO(1.5)-UO(2+x)-ZrO(2) system in air are presented. The eutectic position and the content of the phases crystallized at this point have been determined. The temperature and the composition of the ternary eutectic are 1323 +/- 7 degrees C and 67.4 +/- 1.0 FeO(1.5), 30.5 +/- 1.0 UO(2+x), 2.1 +/- 0.2 ZrO(2) mol.%, respectively. The solubilities of FeO(1.5) and ZrO(2) in the UO(2+x)(FeO(1.5), ZrO(2)) solid solution correspond to respectively 3.2 and 1.1 mol.%. The solubilities of UO(2) and ZrO(2) in FeO(1.5) are not significant. The existence of a solid solution on the basis of U(Zr)FeO(4) compound is found. The ZrO(2) Solubility in this solid solution is 7.0 mol.%.

  • 2.
    Almjashev, V.I.
    et al.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Barrachin, M.
    Institut de Radioprotection et Suˆrete´ Nucle´aire (IRSN).
    Bechta, Sevostian
    DSAR, AP Aleksandrov Res Inst Technol, NITI, Sosnovyi Bor 188540, Russia .
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Defoort, F.
    Laboratoire de Physico-chimie et Thermohydraulique Multiphasiques (LPTM), CEA/Grenoble, DTN/SE2T/LPTM – 17 rue des Martyrs.
    Fischer, M.
    Framatome ANP GmbH.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Hellmann, S.
    Framatome ANP GmbH.
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Lopukh, D.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Mezentseva, L.P.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Miassoedov, A.
    Institute for Nuclear and Energy Technologies, Forschungszentrum Karlsruhe.
    Petrov, Yu.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Vitol, S.A
    A.P. Aleksandrov Research Institute of Technology.
    Phase equilibria in the FeO(1+x)-UO(2)-ZrO(2) system in the FeO(1+x)-enriched domain2010In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 400, no 2, p. 119-126Article in journal (Refereed)
    Abstract [en]

    Experimental results of the investigation of the FeO(1+x)UO(2)-ZrO(2) system in neutral atmosphere are presented. The ternary eutectic position and the composition of the phases crystallized at this point have been determined. The phase diagram is constructed for the FeO(1+x)-enriched region and the onset melting temperature of 1310 degrees C probably represents a local minimum and so will be a determining factor in this system and its application to safety studies in nuclear reactors.

  • 3.
    Almjashev, V.I.
    et al.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Barrachin, M.
    Institut de Radioprotection et Suˆrete´ Nucle´aire (IRSN).
    Bechta, Sevostian
    A.P. Aleksandrov Research Institute of Technology.
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Vitol, S.A
    A.P. Aleksandrov Research Institute of Technology.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Defoort, F.
    Laboratoire de Physico-chimie et Thermohydraulique Multiphasiques (LPTM), CEA/Grenoble, DTN/SE2T/LPTM – 17 rue des Martyrs.
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Lopukh, D.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Lysenko, A.V.
    Alexandrov Research Institute of Technology, Federal State Unitary Enterprise.
    Martynov, A.P.
    St. Petersburg Electrotechnical University.
    Mezentseva, L.P.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Miassoedov, A.
    Institute for Nuclear and Energy Technologies, Forschungszentrum Karlsruhe.
    Petrov, Yu. B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Fischer, M.
    Framatome ANP GmbH.
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Hellmann, S.
    Framatome ANP GmbH.
    Ternary eutectics in the systems FeO-UO2-ZrO2 and Fe2O3-U3O8-ZrO212011In: Radiochemistry, ISSN 1066-3622, Vol. 53, no 1, p. 13-18Article in journal (Refereed)
    Abstract [en]

    The systems FeO–UO2–ZrO2 (in inert atmosphere) and Fe2O3–U3O8–ZrO2 (in air) were studied. Forthe FeO–UO2–ZrO2 system, the eutectic temperature was found to be 1310°С, with the following componentconcentrations (mol %): 91.8 FeO, 3.8 UO2, and 4.4 ZrO2. For the Fe2O3–U3O8–ZrO2 system, the eutectictemperature was found to be 1323°С, with the following component concentrations (mol %): 67.4 FeO1.5,30.5 UO2.67, and 2.1 ZrO2. The solubility limits of iron oxides in the phases based on UO2(ZrO2,FeO) andUO2.67(ZrO2,FeO1.5) were determined

  • 4. Almyashev, V. I.
    et al.
    Granovsky, V. S.
    Khabensky, V. B.
    Krushinov, E. V.
    Sulatsky, A. A.
    Vitol, S. A.
    Gusarov, V. V.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Barrachin, M.
    Fichot, F.
    Bottomley, P. D.
    Fischer, M.
    Piluso, P.
    Oxidation effects during corium melt in-vessel retention2016In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 305, p. 389-399Article in journal (Refereed)
    Abstract [en]

    In the in-vessel corium retention studies conducted on the Rasplav-3 test facility within the ISTC METCOR-P project and OECD MASCA program, experiments were made to investigate transient processes taking place during the oxidation of prototypic molten corium. Qualitative and quantitative data have been produced on the sensitivity of melt oxidation rate to the type of oxidant, melt composition, molten pool surface characteristics. The oxidation rate is a governing factor for additional heat generation and hydrogen release; also for the time of secondary inversion of oxidic and metallic layers of corium molten pool.

  • 5.
    Asmolov, V.G.
    et al.
    Russian Research Center Kurchatov Institute (RRC KI).
    Bechta, Sevostian
    Alexandrov Research Institute of Technology (NITI), SosnovyBor.
    Berkovich, V.M.
    Moscow Research and Design Institute Atomenergoproekt (AEP).
    Bezlepkin, V.V.
    Saint Petersburg Research and Design Institute - Atomenergoproekt (SPb AEP), Saint Petersburg.
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Kisselev, A.E.
    Nuclear Safety Institute, Russian Academy of Sciences (IBRAE RAS).
    Kukhtevich, I.V.
    Alexandrov Research Institute of Technology (NITI).
    Sidorov, A.S.
    Design Branch of Rosenergoatom (DB REA).
    Strizhov, V.F.
    Nuclear Safety Institute, Russian Academy of Sciences (IBRAE RAS).
    Svetlov, S.V.
    Saint Petersburg Research and Design Institute - Atomenergoproekt (SPb AEP).
    Zagryazkin, V.M.
    Russian Research Center Kurchatov Institute (RRC KI).
    Crucible-type core catcher for VVER-1000 reactor2005In: Proceedings of the American Nuclear Society - International Congress on Advances in Nuclear Power Plants 2005, ICAPP'05, Curran Associates, Inc. , 2005, Vol. 2, p. 1221-1227Conference paper (Refereed)
    Abstract [en]

    For new designs of NPPs with VVER-1000 reactors a crucible-type core catcher has been developed to stabilize and cool down molten corium in the reactor pit. The paper addresses distinguishing features of the concept including the "sacrificial" material and the core catcher design. Main phenomena in the catcher have been analyzed.

  • 6.
    Asmolov, V.G.
    et al.
    Russian Research Center Kurchatov Institute (RRC KI).
    Bechta, Sevostian
    A.P. Alexandrov Research Institute of Technology (NITI).
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Gusarov, V.V
    3 Institute of Silicate Chemistry of Russian Academy of Science (ISC of RAS), St. Petersburg .
    Vishnevsky, V.Yu
    Luch Scientific Production Association, Podolsk, Russi.
    Degaltsev, Yu.A
    Russian Research Centre “Kourchatov Institute” (RRC KI), Moscow.
    Abalin, S.S
    Russian Research Centre “Kourchatov Institute” (RRC KI), Moscow.
    Krushinov, E.V.
    A.P. Alexandrov Research Institute of Technology (NITI), Sosnovy Bor.
    Vitol, S.A.
    A.P. Alexandrov Research Institute of Technology (NITI), Sosnovy Bo.
    Almjashev, V.I.
    Institute of Silicate Chemistry of Russian Academy of Science (ISC of RAS), St. Petersburg .
    Kotova, S.Yu
    A.P. Alexandrov Research Institute of Technology (NITI), Sosnovy Bor.
    Zagryazkin, V.N
    Russian Research Centre “Kourchatov Institute” (RRC KI), Moscow.
    Dyakov, E.K
    Luch Scientific Production Association, Podolsk, Russia.
    Strizhov, V.F.
    Nuclear Safety Institute (IBRAE) Russian Academy of Sciences, Moscow.
    Kiselev, N.P
    Russian Research Centre “Kourchatov Institute” (RRC KI), Moscow.
    Partitioning of Zr, U and FP between Molten Oxidic and Metallic Corium2004In: Proceeding of MASCA Seminar, 2004Conference paper (Refereed)
    Abstract [en]

    Interaction of molten corium and liquid iron/stainless steel has been studied in different tests of theMASCA-1 program. These tests utilized the technology of induction melting in a cold crucible. Themasses of tested corium were approximately 0.5, 2 and 100 kg. Representative quantities of Mo, Ru,SrO, BaO, CeO2 and La2O3 served as fission product simulants.After the suboxidized melt - steel interaction U and Zr have been found in the metallic phase.To quantify the partitioning of Zr, U and fission products an extensive experimental program has beenperformed. The following key parameters have been identified: oxygen potential in the melt (degree ofZr-oxidation), the corium/steel mass ratio and U/Zr ratio. The paper discusses the influence of theseparameters on the partitioning of the main species.

  • 7.
    Asmolov, V.G.
    et al.
    Russian Research Center Kurchatov Institute (RRC KI).
    Sulatsky, A.A.
    Alexandrov Scientific-Research Institute of Technology.
    Bechta, Sevostian
    Aleksandrov Research Institute of Technology.
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Vitol, S.A
    A.P. Aleksandrov Research Institute of Technology.
    Almjashev, V.I.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Strizhov, V.F.
    Nuclear Safety Institute, Russian Academy of Sciences (IBRAE RAS).
    The interaction of nuclear reactor core melt with oxide sacrificial material of localization device for a nuclear power plant with water-moderated water-cooled power reactor2007In: High Temperature, ISSN 0018-151X, E-ISSN 1608-3156, Vol. 45, no 1, p. 22-31Article in journal (Refereed)
    Abstract [en]

    The basic results are given of an experimental investigation of the interaction oxide corium melt containing unoxidized zirconium with the sacrificial material of the device for localization of the core melt of a water-moderated water-cooled power reactor (VVER). The mechanism is determined and a model developed of interaction between suboxidized corium melt and sacrificial material.

  • 8.
    Asmolov, V.G.
    et al.
    Russian Research Center Kurchatov Institute (RRC KI).
    Tsurikov, D.F.
    Bechta, Sevostian
    A.P. Alexandrov Research Institute of Technology (NITI).
    Molten Corium Stratification and Component Partitioning2007In: Proceedings of the MASCA2 Seminar 2007, 2007Conference paper (Refereed)
  • 9. Bakardjieva, S.
    et al.
    Barrachin, M.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. Aleksandrov RIT/NITI, Russia.
    Bezdicka, P.
    Bottomley, D.
    Brissonneau, L.
    Cheynet, B.
    Dugne, O.
    Fischer, E.
    Fischer, M.
    Gusarov, V.
    Journeau, C.
    Khabensky, V.
    Kiselova, M.
    Manara, D.
    Piluso, P.
    Sheindlin, M.
    Tyrpekl, V.
    Wiss, T.
    Quality improvements of thermodynamic data applied to corium interactions for severe accident modelling in SARNET22014In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 74, p. 110-124Article in journal (Refereed)
    Abstract [en]

    In a severe accident transient, corium composition and its properties determine its behaviour and its potential interactions both with the reactor vessel and in the later phases with the concrete basemat. This, in turn, requires a detailed knowledge of the phases present at temperature and how they are formed. Because it implies mainly the investigation of chemical systems at high temperature, these data are often difficult to obtain or are uncertain if it already exists. Therefore more data are required both to complete the thermodynamic databanks (such as NUCLEA) and to construct accurate equilibrium phase diagrams and to finally contribute to the improvement of the codes simulating these severe accident conditions. The MCCI work package (WP6) of the SARNET 2 Network of Excellence has been addressing these problems. In this framework in large facilities such as VULCANO tests have been performed on the interactions and ablation of UO2-containing melts with concrete. They have been completed by large scale MCCI testing such EPICOR on vessel steel corrosion. In parallel in major EU-funded ISTC projects co-ordinated with national institutes, such as the CORPHAD and PRECOS, smaller, single effect tests have been carried out on the more difficult phase diagrams. These have produced data that can be directly used by databanks and for modelling improvement/validation. From these data significant advances in the melt chemistry and pool behaviour have been made. A selection of experiments from participating institutes are presented in this paper and give hindsight into the major processes and so give clear indications for the future work, especially in light of the Fukushima accident.

  • 10.
    Bakardjieva, S.
    et al.
    Institute of Inorganic Chemistry, Czech Acad. Sci., Rez, Czech Republic.
    Barrachin, M.
    Institut de Radioprotection et Suˆrete´ Nucle´aire (IRSN).
    Bechta, Sevostian
    A.P. Alexandrov Research Institute of Technology (NITI).
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Brissoneau, L.
    CEA, DEN, Cadarache, F-13108 St Paul lez Durance, France.
    Cheynet, B.
    Thermodata, 6 rue du Tour de l'Eau, F-38400 St Martin d'Hères, France.
    Fischer, E.
    Thermodata, 6 rue du Tour de l'Eau, F-38400 St Martin d'Hères, France.
    Journeau, C.
    CEA, DEN, Cadarache.
    Kiselova, M.
    Nuclear Research Institute UJV, Rez, 250 68, Czech Republic.
    Mezentseva, L.P.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Piluso, P.
    CEA/DEN/DSNI, Bbt. 121, 91191 Gif sur Yvette Cedex, Saclay, France.
    Wiss, T.
    European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe, Germany.
    Improvement of the European thermodynamic database NUCLEA2010In: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 52, no 1, p. 84-96Article in journal (Refereed)
    Abstract [en]

    Modelling of corium behaviour during a severe accident requires knowledge of the phases present at equilibrium for a given corium composition, temperature and pressure. The thermodynamic database NUCLEA in combination with a Gibbs Energy minimizer is the European reference tool to achieve this goal. This database has been improved thanks to the analysis of bibliographical data and to EU-funded experiments performed within the SARNET network, PLINIUS as well as the ISTC CORPHAD and EVAN projects. To assess the uncertainty range associated with Energy Dispersive X-ray analyses, a round-robin exercise has been launched in which a UO2-containing corium-concrete interaction sample from VULCANO has been analyzed by three European laboratories with satisfactorily small differences.

  • 11.
    Bakardjieva, S.
    et al.
    Institute of Inorganic Chemistry, Czech Acad.Sci.,Rez (CZ).
    Barrachin, M.
    Institut de Radioprotection et Suˆrete´ Nucle´aire (IRSN).
    Bechta, Sevostian
    A.P. Alexandrov Research Institute of Technology (NITI).
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Brissoneau, L.
    CEA, DEN, Cadarache, F-13108 St Paul lez Durance, France.
    Cheynet, B.
    Thermodata, 6 rue du Tour de l'Eau, F-38400 St Martin d'Hères, France.
    Fischer, M.
    Framatome Atomic Nuclear Plant, Erlangen, Germany.
    Journeau, C.
    CEA, DEN, Cadarache.
    Kiselova, M.
    Nuclear Research Institute UJV, Rez, 250 68, Czech Republic.
    Mezentseva, L.P.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Piluso, P.
    CEA/DEN/DSNI, Bbt. 121, 91191 Gif sur Yvette Cedex, Saclay, France.
    Wiss, T.
    European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe, Germany.
    Improvement of the European thermodynamic database NUCLEA in the frame of EU-funded experiments2008In: Proceedings of the 3rd European Review Meeting on Severe Accident Research (ERMSAR 2008), 2008Conference paper (Refereed)
    Abstract [en]

    Modelling of corium behaviour during a severe accident requires knowledge of the phases present at equilibrium for a given corium composition, temperature and pressure. The thermodynamic database NUCLEA in combination with a Gibbs Energy minimizer is theEuropean reference tool to achieve this goal. Within SARNET, this database has beenimproved thanks to the analysis of bibliographical data and to EU0funded experimentsperformed within the SARNET network, PLINIUS as well as the ISTC CORPHAD and EVAN projects. To assess the uncertainty range associated with Energy Dispersive X-ray analyses, a round robin exercise has been launched in which a UO2-containing coriumconcrete interaction sample from VULCANO has been analyzed by three European laboratories with satisfactorily small differences.

  • 12.
    Bechta, Sevostian
    et al.
    A.P. Alexandrov Research Institute of Technology (NITI).
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Almjashev, V.I.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Mezentseva, L.P.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Kotova, S.Yu.
    Alexandrov Research Institute of Technologies (NITI).
    Kosarevsky, R.A.
    Alexandrov Research Institute of Technologies (NITI).
    Barrachin, M.
    Institut de Radioprotection et Suˆrete´ Nucle´aire (IRSN).
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Fichot, F.
    Institut de Radioprotection et Sûreté Nucléaire IRSN/DPAM.
    Fischer, M.
    Framatome ANP GmbH.
    Corium phase equilibria based on MASCA, METCOR and CORPHAD results2008In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 238, no 10, p. 2761-2771Article in journal (Refereed)
    Abstract [en]

    Experimental data on component partitioning between suboxidized corium melt and steel in the invessel melt retention (IVR) conditions are compared. The data are produced within the OECD MASCAprogram and the ISTC CORPHAD project under close-to-isothermal conditions and in the ISTC METCORproject under thermal gradient conditions. Chemical equilibrium in the U–Zr–Fe(Cr,Ni,. . .)–O system isreached in all experiments. In MASCA tests the molten pool formed under inert atmosphere has twoimmiscible liquids, oxygen-enriched (oxidic) and oxygen-depleted (metallic), resulting of the miscibilitygap of the mentioned system. Sub-system data of the U–Zr–Fe(Cr,Ni,. . .)–O phase diagram investigatedwithin the ISTC CORPHAD project are interpreted in relation with the MASCA results. In METCOR teststhe equilibrium is established between oxidic liquid and mushy metallic part of the system. Results ofcomparison are discussed and the implications for IVR noted.

  • 13.
    Bechta, Sevostian
    et al.
    Aleksandrov Research Institute of Technology, NITI, Sosnovy Bor.
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Vitol, S.A
    A.P. Aleksandrov Research Institute of Technology.
    Strizhov, V.F.
    Nuclear Safety Institute, Russian Academy of Sciences (IBRAE RAS).
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Fischer, M.
    Framatome Atomic Nuclear Plant, Erlangen, Germany.
    Piluso, P.
    CEA/DEN/DSNI, Bbt. 121, 91191 Gif sur Yvette Cedex, Saclay, France.
    Miassoedov, A.
    Institute for Nuclear and Energy Technologies, Forschungszentrum Karlsruhe.
    Tromm, W.
    FZK, IKET, Karlsruhe, Forschungszentrum Karlsruhe GmbH.
    Altstadt, E.
    Forschungszentrum Rossendorf (FZR), 01328 Dresden, Germany.
    Willschutz, H.G.
    Forschungszentrum Dresden, FZD, Dresden (DE).
    Fichot, F.
    Institut de Radioprotection et Sûreté Nucléaire IRSN/DPAM.
    Kymalainen, O.
    FORTUM Nuclear services Ltd, Espoo (FI).
    VVER steel corrosion during in-vessel retention of corium melt2008In: Proceedings of the 3rd European Review Meeting on Severe Accident Research (ERMSAR 2008), 2008Conference paper (Refereed)
    Abstract [en]

    Physicochemical phenomena taking place at the corium-steel interaction during theexternal cooling of reactor vessel can result in high-temperature steel corrosion and thinningof the vessel wall. The ISTC METCOR project's experimental studies have shown that themain factors influencing corrosion depth and kinetics are oxygen potential, melt compositionand steel interfacial temperature but also melt – vessel heat flux.Experimental data are used for building a model for VVER vessel steel corrosion undercorium thermochemical loads and for correlations to quantitatively analyze the influence ofcorrosion on the rector vessel thinning. The finite-element calculations, in which thedeveloped models of corrosion and heat transfer in corium pool were used, were able toreproduce the temperature and stress-and-strain vessel condition.

  • 14.
    Bechta, Sevostian
    et al.
    Alexandrov Scientific-Research Technology Institute, Sosnovy Bor.
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Vitol, S.A
    A.P. Aleksandrov Research Institute of Technology.
    Sulatsky, A.A.
    Alexandrov Scientific-Research Institute of Technology.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Almjashev, V.I.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Lopukh, D.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Fischer, M.
    Framatome ANP GmbH.
    Piluso, P.
    CEA/DEN/DSNI, Bbt. 121, 91191 Gif sur Yvette Cedex, Saclay, France.
    Miassoedov, A.
    Institute for Nuclear and Energy Technologies, Forschungszentrum Karlsruhe.
    Tromm, W.
    FZK, IKET, Karlsruhe, Forschungszentrum Karlsruhe GmbH.
    Altstadt, E.
    Forschungszentrum Rossendorf (FZR), 01328 Dresden, Germany.
    Fichot, F.
    Institut de Radioprotection et Sûreté Nucléaire IRSN/DPAM.
    Kymaelaeinen, O.
    Fortum Nuclear Services Ltd, POB 10, FIN-00048 Fortum (Finland).
    INTERACTION BETWEEN MOLTEN CORIUM UO2+x-ZrO2-FeOy AND VVER VESSEL STEEL2010In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 170, no 1, p. 210-218Article in journal (Refereed)
    Abstract [en]

    In case of in-vessel corium retention during a severe accident in a light water reactor, weakening of the vessel wall and deterioration of the vessel steel properties can be caused both by the melting of the steel and by its physicochemical interaction with corium. The interaction behavior has been studied in medium-scale experiments with prototypic corium. The experiments yielded data for the steel corrosion rate during interaction with UO2+x-ZrO2-FeOy melt in air and steam at different steel surface temperatures and heat fluxes from the corium to the steel. It has been observed that the corrosion rates in air and steam atmosphere are almost the same. Further, if the temperature at the interface increases beyond a certain level, corrosion intensifies. This is explained by the formation of liquid phases in the interaction zone. The available experimental data have been used to develop a correlation for the corrosion rate as afunction of temperature and heat flux.

  • 15.
    Bechta, Sevostian
    et al.
    Alexandrov Scientific-Research Technology Institute, Sosnovy Bor.
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Vitol, S.A
    A.P. Aleksandrov Research Institute of Technology.
    Sulatsky, A.A.
    Alexandrov Scientific-Research Institute of Technology.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Almjashev, V.I.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Lopukh, D.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Fischer, M.
    Framatome ANP GmbH.
    Piluso, P.
    CEA/DEN/DSNI, Bbt. 121, 91191 Gif sur Yvette Cedex, Saclay, France.
    Miassoedov, A.
    Institute for Nuclear and Energy Technologies, Forschungszentrum Karlsruhe.
    Tromm, W.
    FZK, IKET, Karlsruhe, Forschungszentrum Karlsruhe GmbH.
    Altstadt, E.
    Forschungszentrum Rossendorf (FZR), 01328 Dresden, Germany.
    Fichot, F.
    Institut de Radioprotection et Sûreté Nucléaire IRSN/DPAM.
    Kymalainerf, O.
    FORTUM Nuclear Services Ltd, Espoo, Finland.
    Interaction between molten corium UO2+X-ZrO2-FeO y and VVER vessel steel2009In: Proceeding of International Conference on Advances in Nuclear Power Plants, ICAPP 2008, Curran Associates, Inc., 2009, Vol. 170, p. 210-218Conference paper (Refereed)
    Abstract [en]

    In case of an in-vessel corium retention (1VR) the deterioration of vessel steel properties can be caused both by the steel melting and by its physicochemical interaction with corium. The interaction behavior has been studied in the medium-scale experiments with a prototypic corium within the METCOR project. The resulting experimental data give an insight into the steel corrosion during its interaction with U02+x- Zr02- FeOy melt in air and steam. It has been observed that the corrosion rate is almost the same in air and steam atmosphere; if the temperature on the interaction interface increases beyond a certain level, corrosion intensifies, which is explained by the formation of liquid phases in the interaction zone. The available experimental data have been used for developing a correlation of corrosion rate versus temperature and heat flux.

  • 16.
    Bechta, Sevostian
    et al.
    A.P. Aleksandrov Research Institute of Technology.
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Vitol, S.A
    A.P. Aleksandrov Research Institute of Technology.
    Sulatsky, A.A.
    Alexandrov Scientific-Research Institute of Technology.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Almjashev, V.I.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Lopukh, D.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Fischer, M.
    Framatome ANP GmbH.
    Piluso, P.
    CEA/DEN/DSNI, Bbt. 121, 91191 Gif sur Yvette Cedex, Saclay, France.
    Miassoedov, A.
    Institute for Nuclear and Energy Technologies, Forschungszentrum Karlsruhe.
    Tromm, W.
    FZK, IKET, Karlsruhe, Forschungszentrum Karlsruhe GmbH.
    Altstadt, E.
    Forschungszentrum Rossendorf (FZR), 01328 Dresden, Germany.
    Fichot, F.
    Institut de Radioprotection et Sûreté Nucléaire IRSN/DPAM.
    Kymalainerf, O.
    FORTUM Nuclear Services Ltd, Espoo, Finland.
    VVER vessel steel corrosion at interaction with molten corium in oxidizing atmosphere2009In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 239, no 6, p. 1103-1112Article in journal (Refereed)
    Abstract [en]

    The long-term in-vessel corium retention (IVR) in the lower head bears a risk of the vessel wall deterioration caused by steel corrosion. The ISTC METCOR Project has studied physicochemical impact of prototypic coria having different compositions in air and steam and has generated valuable experimental data on vessel steel corrosion. It is found that the corrosion rate is sensitive to corium composition, but the composition of oxidizing above-melt atmosphere (air, steam) has practically no influence on it. A model of the corrosion process that integrates the experimental data, is proposed and used for development of correlations.

  • 17.
    Bechta, Sevostian
    et al.
    A.P. Alexandrov Research Institute of Technology (NITI).
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    Sulatsky, A.A.
    Alexandrov Scientific-Research Institute of Technology.
    Model of Interaction of Molten Steel with Sacrificial Material of VVER Core Catcher2004In: Proceedings of Research Workshop “Evaluation of experimental data and verification of computer codes", Russia, 2004Conference paper (Refereed)
  • 18.
    Bechta, Sevostian
    et al.
    Scientific Research Technological Institute (NITI), Russian Federation.
    Khabensky, V. B.
    Vitol, S. A.
    Krushinov, E. V.
    Lopukh, D. B.
    Petrov, Yu.B.
    Petchenkov, A.Yu.
    Kulagin, I. V.
    Granovsky, V. S.
    Kovtunova, S. V.
    Martinov, V. V.
    Gusarov, V. V.
    Experimental studies of oxidic molten corium-vessel steel interaction2001In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 210, no 1-3, p. 193-224Article in journal (Refereed)
    Abstract [en]

    The experimental results of molten corium-steel specimen interaction with molten corium on the 'Rasplav-2' test facility are presented. In the experiments, cooled vessel steel specimens positioned on the molten pool bottom and uncooled ones lowered into the molten pool were tested. Interaction processes were studied for different corium compositions, melt superheating and in alternative (inert and air) overlying atmosphere. Hypotheses were put forward explaining the observed phenomena and interaction mechanisms. The studies presented in the paper were aimed at the detection of different corium-steel interaction mechanisms. Therefore certain identified phenomena are more typical of the ex-vessel localization conditions than of the in-vessel corium retention. Primarily, this can be referred to the phenomena of low-temperature molten corium-vessel steel interaction in oxidizing atmosphere.

  • 19.
    Bechta, Sevostian
    et al.
    A.P. Alexandrov Research Institute of Technology (NITI).
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    ISTC CORPHAD Project: Experimental study of corium phase diagram2006In: Proceedings of International Information Exchange Meeting on Thermodynamics of Nuclear Fuels, 2006Conference paper (Refereed)
  • 20.
    Bechta, Sevostian
    et al.
    A.P. Alexandrov Research Institute of Technology (NITI).
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Vitol, S.A
    A.P. Aleksandrov Research Institute of Technology.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Almjashev, V.I.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Lopukh, D.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Tromm, W.
    FZK, IKET, Karlsruhe, Forschungszentrum Karlsruhe GmbH.
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Fischer, M.
    Framatome ANP GmbH.
    Cognet, G.
    CEA/DEN/DSNI, Bbt 121, 91191 Gif sur Yvette Cedex, Saclay (France).
    Kymäläinen, O.
    Fortum Nuclear Services Ltd, POB 10, FIN-00048 Fortum, Finland.
    New experimental results on the interaction of molten corium with reactor vessel steel2004In: Proceedings of the 2004 International Congress on Advances in Nuclear Power Plants, ICAPP'04, American Nuclear Society, 2004, p. 1072-1081Conference paper (Refereed)
    Abstract [en]

    In order to justify the concept of in-vessel core melt retention, it is necessary to understand the thermal and physico-chemical phenomena. Especially the interaction of the molten pool with the reactor vessel during outside cooling needs to be understood. These phenomena are very complex, in particular, where interactions with the oxidic melt are concerned. In the early stages of the retention process, the oxidic corium and the vessel steel interact under the conditions of low oxygen potential in the melt. These conditions can be simulated by a molten corium having the composition UO2/ZrO 2Zr, where the degree of Zr-oxidation is in the range between 30 % (C-30) and 100 % (C-100). Corresponding experiments with prototypic melts at low oxygen potentials are being performed in the ISTC METCOR project 2nd phase. These are: MC 5 of corium composition 71w%UO2-29w%ZrO 2 (C-100) in neutral atmosphere (argon), MC 6 of corium composition 76w%UO2-9w%ZrO2-15w%Zr (C∼30), also in argon. In test MC 5, the interaction of molten C-100 corium with a water-cooled steel specimen was studied for the following maximum temperatures at the specimen surface: 1075°C, 1180°C, 1315°C and 1435°C. The total duration of the experiment was ∼ 36 hours. The MC5 test serves as a reference test for determining the characteristics of the interaction between oxidic melt and steel specimen under the conditions of minimum chemical interaction potential. To investigate the effect of substoichiometry, test MC 6 was then performed with suboxidized molten corium C∼30. The maximum surface temperature of the cooled steel specimen was held at ∼ 1400°C. The test duration was ∼ 10 hours. The ablation phenomena were found to differ significantly from those observed both in the reference test, as well as in former tests with oxidized melts, as they involved the formation of a low-melting metallic phase at the interface which contains iron, zirconium and uranium. The paper summarizes the results of the experiments and of the performed posttest analysis for tests MC 5 and MC 6.

  • 21.
    Bechta, Sevostian
    et al.
    Alexandrov Research Institute of Technology (NITI).
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Vitol, S.A
    A.P. Aleksandrov Research Institute of Technology.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Almjashev, V.I.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Lopukh, D.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Tromm, W.
    FZK, IKET, Karlsruhe, Forschungszentrum Karlsruhe GmbH.
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Fischer, M.
    Framatome ANP GmbH.
    Piluso, P.
    CEA/DEN/DSNI, Bbt. 121, 91191 Gif sur Yvette Cedex, Saclay, France.
    Miassoedov, A.
    Institute for Nuclear and Energy Technologies, Forschungszentrum Karlsruhe.
    Altstadt, E.
    Forschungszentrum Rossendorf (FZR), 01328 Dresden, Germany.
    Willschütz, H.G.
    Forschungszentrum Rossendorf (FZR), 01328 Dresden, Germany.
    Fichot, F.
    Institut de Radioprotection et Sûreté Nucléaire IRSN/DPAM.
    Experimental study of interactions between suboxidized corium and reactor vessel steel2006In: Proceedings of the 2006 International Congress on Advances in Nuclear Power Plants, ICAPP'06, 2006, Vol. 2006, p. 1355-1362Conference paper (Refereed)
    Abstract [en]

    One of the critical factors in the analysis of in-vessel melt retention is the vessel strength. It is, in particular, sensitive to the thickness of intact vessel wall, which, in its turn, depends on the thermal conditions and physicochemical interactions with corium. Physicochemical interaction of prototypic UO2-ZrO2-Zr corium melt and VVER vessel steel was examined during the 2nd Phase of the ISTC METCOR Project. Rasplav-3 test facility was used for conducting four tests, in which the Zr oxidation degree and interaction front temperature were varied; in one of the tests, stainless steel was added to the melt. Direct experimental measurements and posttest analyses were used for determining corrosion kinetics and maximum corrosion depth (i.e. the physicochemical impact of corium on the cooled vessel steel specimens), as well as the steel temperature conditions during the interaction, and finally the structure and composition of crystallized ingots, including the interaction zone. The minimum temperature on the interaction front boundary, which determined its final position and maximum corrosion depth was ∼ 1090°C. An empirical correlation for calculation of corrosion kinetics has been derived.

  • 22.
    Bechta, Sevostian
    et al.
    A.P. Alexandrov Research Institute of Technology (NITI).
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Vitol, S.A
    A.P. Aleksandrov Research Institute of Technology.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Almjashev, V.I.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Mezentseva, L.P.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Petrov, Yu.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Lopukh, D.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Fischer, M.
    Framatome Atomic Nuclear Plant, Erlangen, Germany.
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Tromm, W.
    FZK, IKET, Karlsruhe, Forschungszentrum Karlsruhe GmbH.
    Barrachin, M.
    Institut de Radioprotection et Suˆrete´ Nucle´aire (IRSN).
    Altstadt, E.
    Forschungszentrum Rossendorf (FZR), 01328 Dresden, Germany.
    Piluso, P.
    CEA/DEN/DSNI, Bbt. 121, 91191 Gif sur Yvette Cedex, Saclay, France.
    Fichot, F.
    Institut de Radioprotection et Sûreté Nucléaire IRSN/DPAM.
    Hellmann, S.
    Framatome ANP GmbH.
    Defoort, F.
    Laboratoire de Physico-chimie et Thermohydraulique Multiphasiques (LPTM), CEA/Grenoble, DTN/SE2T/LPTM – 17 rue des Martyrs.
    CORPHAD and METCOR ISTC projects2005In: Proceedings of The first European Review Meeting on Severe Accident Research (ERMSAR-2005), 2005Conference paper (Refereed)
    Abstract [en]

    The ongoing CORPHAD Project (Phase Diagrams for Multicomponent SystemsContaining Corium and Products of its Interaction with NPP Materials) started in August2001. The main aim of the project is to experimentally determine the relevantphysicochemical data on phase diagrams of binary, ternary, quaternary and prototypic multicomponent systems, which are important for analysis and modelling of a severe accident (SA)and efficient planning of severe accident management (SAM) measures. The data should bedirectly used for the European NUCLEA database development and validation. The followingsystems are in the focus of the project: (1) UO2 – FeO, (2) ZrO2 – FeO, (3) SiO2– Fe2O3, (4)UO2 – SiO2, (5) UO2 – ZrO2-FeO, (6) UO2 – ZrO2-FeOy, (7) U-O-Fe, (8) Zr-O-Fe, (9) U-OZr, (10) U-Zr-Fe-O, (11) complex corium mixtures.The experimentally determined data of the listed diagrams include: coordinates ofcharacteristic points (eutectics, peritectics and others); liquidus and solidus concentrationcurves; component solubility limits in the solid phase; tie line coordinates and temperatureconcentration regions of the miscibility gap. Different methodologies are used for the phasediagram study. Classical methods of thermal analysis, like DTA and DSC are combined withmethods specifically developed for corium studies.The METCOR project (Investigation of Corium Melt Interaction with NPP ReactorVessel Steel) started in April 1999. The objectives of the project are to qualify and to quantifyphysico-chemical phenomena of corium melt interaction with reactor vessel steel cooled fromthe outside. The variable parameters of the interaction tests are: oxygen potential in thesystem, corium composition, interaction interface temperature and heat flux from corium tosteel. The medium scale tests with corium mass of about 2 kg are carried out by using highfrequency induction heating of the corium melt in a cold crucible.The METCOR & CORPHAD work-packages are performed by Russian partners inclose collaboration with leading European scientific institutes in the area of corium researchas well as with the European nuclear industry.This paper briefly describes the results obtained in both projects and their possibleapplication for SA analysis and SAM. The paper concludes with recommendations for futureresearch activities in the framework of METCOR and CORPHAD projects.

  • 23.
    Bechta, Sevostian
    et al.
    A.P. Alexandrov Research Institute of Technology (NITI).
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Vitol, S.A
    A.P. Aleksandrov Research Institute of Technology.
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    Lopukh, D.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Martinov, A.P.
    SPb Electrotechnical University (SPbGETU).
    Martinov, A.P.
    SPb Electrotechnical University (SPbGETU).
    Fieg, G.
    Forshungszentrum Karlsruhe (FZK), Institut fur Neutronenphysik and Reaktortechnik.
    Tromm, W.
    FZK, IKET, Karlsruhe, Forschungszentrum Karlsruhe GmbH.
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Tuomisto, H.
    Fortum Engineering Ltd..
    Corrosion of vessel steel during its interaction with molten corium: Part 1. Experimental2006In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 236, no 17, p. 1810-1829Article in journal (Refereed)
    Abstract [en]

    This paper is concerned with corrosion of a cooled vessel steel structure interacting with molten corium in air and neutral (nitrogen) atmospheresduring an in-vessel retention scenario. The data on corrosion kinetics at different temperatures on the heated steel surface, heat flux densities andoxygen potential in the system are presented. The post-test physico-chemical and metallographic analyses of melt samples and the corium–specimeningot have clarified certain mechanisms of steel corrosion taking place during the in-vessel melt interaction.

  • 24.
    Bechta, Sevostian
    et al.
    A.P. Alexandrov Research Institute of Technology (NITI).
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Vitol, S.A
    A.P. Aleksandrov Research Institute of Technology.
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    Lopukh, D.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Martinov, A.P.
    SPb Electrotechnical University (SPbGETU).
    Martinov, A.P.
    SPb Electrotechnical University (SPbGETU).
    Fieg, G.
    Forshungszentrum Karlsruhe (FZK), Institut fur Neutronenphysik and Reaktortechnik.
    Tromm, W.
    FZK, IKET, Karlsruhe, Forschungszentrum Karlsruhe GmbH.
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Tuomisto, H.
    Fortum Engineering Ltd..
    Corrosion of vessel steel during its interaction with molten corium: Part 2. Model development2006In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 236, no 13, p. 1362-1370Article in journal (Refereed)
    Abstract [en]

    An experimental examination of the cooled vessel steel corrosion during the interaction with molten corium is presented. The experiments havebeen conducted on “Rasplav-2” test facility and followed up with physico-chemical and metallographic analyses of melt samples and coriumspecimeningots. The results discussed in the first part of the paper have revealed specific corrosion mechanisms for air and inert atmosphere abovethe melt. Models have been proposed based on this information and approximate curves constructed for the estimation of the corrosion rate orcorrosion depth of vessel steel in conditions simulated by the experiments.

  • 25.
    Bechta, Sevostian
    et al.
    A.P. Alexandrov Research Institute of Technology (NITI).
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Almjashev, V.I.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Vitol, S.A
    A.P. Aleksandrov Research Institute of Technology.
    Mezentseva, L.P.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Petrov, Yu.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Lopukh, D.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Barrachin, M.
    Institut de Radioprotection et Suˆrete´ Nucle´aire (IRSN).
    Hellmann, S.
    Framatome ANP GmbH.
    Froment, K.
    Laboratoire de Physico-chimie et Thermohydraulique Multiphasiques (LPTM), CEA/Grenoble, DTN/SE2T/LPTM – 17 rue des Martyrs.
    Fischer, M.
    Framatome ANP GmbH.
    Tromm, W.
    FZK, IKET, Karlsruhe, Forschungszentrum Karlsruhe GmbH.
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Defoort, F.
    Laboratoire de Physico-chimie et Thermohydraulique Multiphasiques (LPTM), CEA/Grenoble, DTN/SE2T/LPTM – 17 rue des Martyrs.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Phase diagram of the UO2-FeO1+x system2007In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 362, no 1, p. 46-52Article in journal (Refereed)
    Abstract [en]

    Phase-relation studies of the UO2–FeO1+x system in an inert atmosphere are presented. The eutectic point has beendetermined, which corresponds to a temperature of (1335 ± 5) C and a UO2 concentration of (4.0 ± 0.1) mol.%. Themaximum solubility of FeO in UO2 at the eutectic temperature has been estimated as (17.0 ± 1.0) mol.%. Liquidus temperaturesfor a wide concentration range have been determined and a phase diagram of the system has been constructed.

  • 26.
    Bechta, Sevostian
    et al.
    A.P. Alexandrov Research Institute of Technology (NITI).
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Almjashev, V.I.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Vitol, S.A
    A.P. Aleksandrov Research Institute of Technology.
    Mezentseva, L.P.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Petrov, Yu.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Lopukh, D.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Barrachin, M.
    Institut de Radioprotection et Suˆrete´ Nucle´aire (IRSN).
    Hellmann, S.
    Framatome ANP GmbH.
    Froment, K.
    Laboratoire de Physico-chimie et Thermohydraulique Multiphasiques (LPTM), CEA/Grenoble, DTN/SE2T/LPTM – 17 rue des Martyrs.
    Fischer, M.
    Framatome ANP GmbH.
    Tromm, W.
    FZK, IKET, Karlsruhe, Forschungszentrum Karlsruhe GmbH.
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Defoort, F.
    Laboratoire de Physico-chimie et Thermohydraulique Multiphasiques (LPTM), CEA/Grenoble, DTN/SE2T/LPTM – 17 rue des Martyrs.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Phase diagram of the ZrO2-FeO system2006In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 348, no 1-2, p. 114-121Article in journal (Refereed)
    Abstract [en]

    The results on the ZrO2–FeO system studies in a neutral atmosphere are presented. The refined eutectic point has beenfound to correspond to a ZrO2 concentration of 10.3 ± 0.6 mol% at 1332 ± 5 C. The ultimate solubility of iron oxide inzirconia has been determined in a broad temperature range, taking into account the ZrO2 polymorphism. A phase diagramof the pseudobinary system in question has been constructed.

  • 27.
    Bechta, Sevostian
    et al.
    A.P. Aleksandrov Research Institute of Technology.
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Almjashev, V.I.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Vitol, S.A
    A.P. Aleksandrov Research Institute of Technology.
    Mezentseva, L.P.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Petrov, Yu.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Lopukh, D.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Barrachin, M.
    Institut de Radioprotection et Suˆrete´ Nucle´aire (IRSN).
    Hellmann, S.
    Framatome ANP GmbH.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Phase Relations in the ZrO2–FeO System2006In: Russian Journal of Inorganic Chemistry, ISSN 0036-0236, E-ISSN 1531-8613, Vol. 51, no 2, p. 325-331Article in journal (Refereed)
    Abstract [en]

    We present the results of the investigation of the ZrO2-FeO system under an inert atmosphere. We have refined the position of the eutectic point, which lies at 1332 +/- 5 degrees C and 10.3 +/- 0.6 mol % ZrO2. The iron oxide solubility boundaries in zirconium dioxide have been determined over a wide temperature range taking into account the polymorphism in ZrO2. A phase diagram for the system has been designed.

  • 28.
    Bechta, Sevostian
    et al.
    Aleksandrov Research Technological Institute.
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Almjashev, V.I.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Vitol, S.A
    A.P. Aleksandrov Research Institute of Technology.
    Mezentseva, L.P.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Petrov, Yu.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Lopukh, D.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Lomanova, N.A.
    b Grebenshchikov Institute of Silicate Chemistry, Russian Academy of Sciences, St. Petersburg, Russia.
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Barrachin, M.
    Institut de Radioprotection et Suˆrete´ Nucle´aire (IRSN).
    Hellmann, S.
    Framatome ANP GmbH.
    Froment, K.
    Laboratoire de Physico-chimie et Thermohydraulique Multiphasiques (LPTM), CEA/Grenoble, DTN/SE2T/LPTM – 17 rue des Martyrs.
    Fischer, M.
    Framatome Atomic Nuclear Plant, Erlangen, Germany.
    Tromm, W.
    FZK, IKET, Karlsruhe, Forschungszentrum Karlsruhe GmbH.
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Phase transformation in the binary section of the UO2-FeO-Fe system2007In: Radiochemistry (New York, N.Y.), ISSN 1066-3622, E-ISSN 1608-3288, Vol. 49, no 1, p. 20-24Article in journal (Refereed)
    Abstract [en]

    Phase transformations in the oxide binary section of the UO2-FeO-Fe ternary system were studied. The melting onset point of the UO2-FeO heterogeneous system (1335±5°C) was determined and the fusion curve of this system was constructed. The limiting solubility of FeO in the UO2 solid solution was measured. The changes in crystal parameters in formation of the solid solution were determined. Uranium dioxide was found to be insoluble in the wüstite phase (FeO).

  • 29.
    Bechta, Sevostian
    et al.
    A.P. Alexandrov Research Institute of Technology (NITI).
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Vitol, S.A
    A.P. Aleksandrov Research Institute of Technology.
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Kotova, S.Yu.
    Alexandrov Research Institute of Technologies (NITI).
    Sulatsky, A.A.
    Alexandrov Scientific-Research Institute of Technology.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Almyashev, V.I.
    Grebenschikov Institute of Silicate Chemistry of the Russian Academy of Sciences.
    Ducros, G.
    CEA, DEN, Cadarache.
    Journeau, C.
    CEA, DEN, Cadarache.
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Clément, B.
    Institut de Radioprotection et Sûreté Nucléaire.
    Herranz, L.
    CIEMAT.
    Guentay, S.
    PSI.
    Trambauer, K.
    GRS.
    Auvinen, A.
    VTT.
    Bezlepkin, V.V.
    SPbAEP.
    Influence of corium oxidation on fission product release from molten pool2009In: Proceedings of 2009 International Congress on Advances in Nuclear Power Plants: ICAPP 2009, 2009Conference paper (Refereed)
  • 30.
    Bechta, Sevostian
    et al.
    A.P. Alexandrov Research Institute of Technology (NITI).
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Vitol, S.A
    A.P. Aleksandrov Research Institute of Technology.
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Kotova, S.Yu.
    Alexandrov Research Institute of Technologies (NITI).
    Sulatsky, A.A.
    Alexandrov Scientific-Research Institute of Technology.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Almyashev, V.I.
    Grebenschikov Institute of Silicate Chemistry of the Russian Academy of Sciences.
    Ducros, G.
    CEA, DEN, Cadarache.
    Journeau, C.
    CEA, DEN, Cadarache.
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Clément, B.
    Institut de Radioprotection et Sûreté Nucléaire.
    Herranz, L.
    CIEMAT.
    Guentay, S.
    Paul Scherrer Institut (PSI).
    Trambauer, K.
    GemResearch Swisslab (GRS).
    Auvinen, A.
    Technical Research Centre of Finland (VTT).
    Bezlepkin, V.V.
    SPbAEP.
    Influence of corium oxidation on fission product release from molten pool2010In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 240, no 5, p. 1229-1241Article in journal (Refereed)
    Abstract [en]

    Qualitative and quantitative determination of the release of low-volatile fission products and core materialsfrom molten oxidic corium was investigated in the EVAN project under the auspices of ISTC. Theexperiments carried out in a cold crucible with induction heating and RASPLAV test facility are described.The results are discussed in terms of reactor application; in particular, pool configuration, melt oxidationkinetics, critical influence of melt surface temperature and oxidation index on the fission productrelease rate, aerosol particle composition and size distribution. The relevance of measured high releaseof Sr from the molten pool for the reactor application is highlighted. Comparisons of the experimentaldata with those from the COLIMA CA-U3 test and the VERCORS tests, as well as with predictions fromIVTANTHERMO and GEMINI/NUCLEA codes are made. Recommendations for further investigations areproposed following the major observations and discussions.

  • 31.
    Bechta, Sevostian
    et al.
    Sci. Res. Technol. Institute (NITI), Russian Federation.
    Vitol, S. A.
    Krushinov, E. V.
    Granovsky, V. S.
    Sulatsky, A. A.
    Khabensky, V. B.
    Lopukh, D. B.
    Petrov, Y. B.
    Pechenkov, A. Y.
    Water boiling on the corium melt surface under VVER severe accident conditions2000In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 195, no 1, p. 45-56Article in journal (Refereed)
    Abstract [en]

    Experimental results are presented on the interaction of corium melt with water supplied on its surface. The tests were conducted in the `Rasplav-2' experimental facility. Corium melt was generated by induction melting in the cold crucible. The following data were obtained: heat transfer at boiling water-melt surface interaction, gas and aerosol release, post-interaction solidified corium structure. The corium melt charge had the following composition, mass%: 60% UO2+x-16% ZRO2-15% Fe2O3-6% Cr2O3-3% Ni2O3. The melt surface temperature ranged within 1920-1970 K.

  • 32.
    Bechta, Sevostian
    et al.
    A.P. Alexandrov Research Institute of Technology (NITI).
    Vitol, S.A
    Kalinin, B.D.
    Mosichev, V.I.
    Plotnikov, R.I.
    Sergeev, Yu.I.
    Prospects of application of quantitative x-ray fluorescence analysis in nuclear industry: Existing solutions and new approaches2009In: Proceeding of the 5th interindustrial meeting “Problems and developments of chemical and radiochemical monitoring in atomic energy”, 2009Conference paper (Refereed)
  • 33. Bottomley, D.
    et al.
    Stuckert, J.
    Hofmann, P.
    Tocheny, L.
    Hugon, M.
    Journeau, C.
    Clement, B.
    Weber, S.
    Guentay, S.
    Hozer, Z.
    Herranz, L.
    Schumm, A.
    Oriolo, F.
    Altstadt, E.
    Krause, M.
    Fischer, M.
    Khabensky, V. B.
    Bechta, Sevostian V.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Veshchunov, M. S.
    Palagin, A. V.
    Kiselev, A. E.
    Nalivaev, V. I.
    Goryachev, A. V.
    Zhdanov, V.
    Baklanov, V.
    Severe accident research in the core degradation area: An example of effective international cooperation between the European Union (EU) and the Commonwealth of Independent States (CIS) by the International Science and Technology Center2012In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 252, p. 226-241Article in journal (Refereed)
    Abstract [en]

    The International Science and Technology Center (ISTC) was set up in Moscow to support non-proliferation of sensitive knowledge and technologies in biological, chemical and nuclear domains by engaging scientists in peaceful research programmes with a broad international cooperation. The paper has two following objectives: to describe the organization of complex, international, experimental and analytical research of material processes under extreme conditions similar to those of severe accidents in nuclear reactors and, to inform briefly about some results of these studies. The main forms of ISTC activity are Research Projects and Supporting Programs. In the Research Projects informal contact expert groups (CEGs) were set up by ISTC to improve coordination between adjacent projects and to encourage international collaboration. The European Commission was the first to use this. The CEG members - experts from the national institutes and industry - evaluated and managed the projects' scientific results from initial stage of proposal formulation until the final reporting. They were often involved directly in the project's details by joining the Steering Committees of the project. The Contact Expert Group for Severe Accidents and Management (CEG-SAM) is one of these groups, five project groups from this area from the total of 30 funded projects during 10 years of activity are detailed to demonstrate this: (1) QUENCH-VVER from RIAR, Dimitrovgrad and IBRAE, Moscow, and PARAMETER projects (SF1-SF4) from LUCH, Podolsk and IBRAE, Moscow; these concerned a detailed study of bundle quenching from high temperature; (2) Reactor Core Degradation; a modelling project simulating the fuel rod degradation and loss of geometry from IBRAE, Moscow; (3) METCOR projects from NITI, St. Petersburg on the interaction of core melt with reactor vessel steel; (4) INVECOR project, NNE Kurchatov City, Kazakhstan; this is a large-scale facility to examine the vessel steel retention of 60 kg corium during the decay heat; and finally, (5) CORPHAD and PRECOS projects, NITI, St. Petersburg undertook a systematic examination of refractory ceramics relevant to in-vessel and ex-vessel coria, particularly examining poorly characterised, limited data or experimentally difficult systems.

  • 34. Dietrich, P.
    et al.
    Kretzschmar, F.
    Miassoedov, A.
    Class, A.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Coupling of melcor with the pecm for improved modelling of a core melt in the lower plenum2015In: International Conference on Nuclear Engineering, Proceedings, ICONE, JSME , 2015Conference paper (Refereed)
    Abstract [en]

    MELCOR contains a coupling interface based on the MPI-Standard, which enables the communication to other codes such as RELAP5 or GASFLOW. However a detailed knowledge of this coupling interface in MELCOR is necessary to use this possibility. Therefore, at the KIT the software tool DINAMO (Direct Interface for Adding Models) has been developed. This program contains the coupling routines as well as an interface to communicate with other programs. Using DINAMO it is also possible to utilize new or enhanced models for phenomena, which occur during a severe accident in a nuclear power plant, in MELCOR without modification of the MELCOR source code. In the present work MELCOR calculations of experiments in the LIVE-Facility are presented. The LIVE-Facility is used to simulate the behavior of a melt in the lower plenum of a reactor pressure vessel (RPV). For these calculations we coupled MELCOR via DINAMO with the Phase-Change Effective Convectivity Model (PECM), which has been developed at the KTH in Stockholm. Using the PECM it is possible to improve the prediction of a core melt in the lower plenum of a RPV in case of a core melt accident.

  • 35. Dietrich, P.
    et al.
    Kretzschmar, F.
    Miassoedov, A.
    Class, A.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Extension of the MELCOR code for analysis of late in-vessel phase of a severe accident2015In: IYCE 2015 - Proceedings: 2015 5th International Youth Conference on Energy, IEEE conference proceedings, 2015Conference paper (Refereed)
    Abstract [en]

    The simulation of severe accidents in nuclear power plants with system codes is a powerful tool to improve the safety measures to prevent severe accidents. The further development of severe accident codes is part of current research. MELCOR, as the leading nuclear safety code, provides the possibility to be coupled to other codes. A detailed knowledge of this coupling interface is necessary to use this possibility. Therefore, the software tool DINAMO, which contains the coupling routines and an interface to communicate with other programs, was developed. Using DINAMO it is possible to utilize new models for specific phenomena in MELCOR. In the present work the Phase-Change Effective Convectivity Model was coupled using the CFD-software OpenFOAM and DINAMO to MELCOR to improve the prediction of molten core material in the lower plenum of a reactor pressure vessel. The simulation results were compared to the experimental findings of the LIVE-facility.

  • 36. Fichot, F.
    et al.
    Carénini, L.
    Sangiorgi, M.
    Hermsmeyer, S.
    Miassoedov, A.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Zdarek, J.
    Guenadou, D.
    Some considerations to improve the methodology to assess In-Vessel Retention strategy for high-power reactors2018In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 119, p. 36-45Article in journal (Refereed)
    Abstract [en]

    The In-Vessel Retention (IVR) strategy for Light Water Reactors (LWR) intends to stabilize and isolate corium and fission products in the reactor pressure vessel and in the primary circuit. This type of Severe Accident Management (SAM) strategy has already been incorporated in the SAM guidance (SAMG) of several operating small size LWR (reactor below 500 MWe (like VVER440)) and is part of the SAMG strategies for some Gen III + PWRs of higher power like the AP1000 or the APR1400. However, for high power reactors, estimations using current level of conservatism show that RPV failure caused by thermo-mechanical rupture takes place in some cases. A better estimation of the residual risk (probability of cases with vessel rupture) requires the use of models with a lower level of conservatism. In Europe, the IVMR project aims at providing new experimental data and a harmonized methodology for IVR. A synthesis of the methodology applied to demonstrate the efficiency of IVR strategy for VVER-440 in Europe (Finland, Slovakia, Hungary and Czech Republic) was made. It showed very consistent results, following quite comparable methodologies. The main weakness of the demonstration was identified in the evaluation of the heat flux that could be reached in transient situations, e.g. under the “3-layers” configuration, where the “focusing effect” may cause higher heat fluxes than in steady-state (due to transient “thin” metal layer on top). Analyses of various designs of reactors with a power between 900 and 1300 MWe were also made. Different models for the description of the molten pool were used: homogeneous, stratified with fixed configuration, stratified with evolving configuration. The last type of model provides the highest heat fluxes (above 3 MW/m2) whereas the first type provides the lowest heat fluxes (around 500 MW/m2) but is not realistic due to the non-miscibility of steel with UO2. Obviously, there is a need to reach a consensus about best estimate practices for IVR assessment to be used in the major codes for safety analysis, such as ASTEC, MELCOR, SOCRAT, MAAP, ATHLET-CD, SCDAP/RELAP, etc. Despite the model discrepancies, and leaving aside the unrealistic case of homogeneous pool, the average calculated heat fluxes in many cases are well above 1 MW/m2 which could reduce the residual thickness of the vessel considerably and threaten its integrity. Therefore, it is clear that the safety demonstration of IVR for high power reactors requires a more careful evaluation of the situations which can lead to formation of either a very thin top metal layer provoking focusing effect or significantly overheated metal, e.g. after oxide and metal layer inversion. It also requires an accurate mechanical analysis of the ablated vessel. The current approach followed by most experts for IVR is a compromise between a deterministic analysis using the significant knowledge gained during the last two decades and a probabilistic analysis to take into account large uncertainties due to the lack of data for some physical phenomena (such as transient effects) and due to excessive simplifications of models. A harmonization of the positions of safety authorities on the IVR strategy is necessary to allow decision making based on shared scientific knowledge. Currently, the acceptance criteria of a safety demonstration for IVR may be differently defined from one country to the other and the differences should be further discussed to reach harmonization on this important topic. This includes the accident scenarios to be considered in the demonstration and the modelling of the phenomena in the vessel. Such harmonization is one of the goals of IVMR project. A revised methodology is proposed, where the safety criterion is based not only on a comparison of the heat flux and the Critical Heat Flux (CHF) profiles as in current approaches but also on the minimum vessel thickness reached after ablation and the maximum integral loads that is applied to the vessel during the transient. The main advantage of this revised criterion is in consideration of both steady-state and transient loads on the RPV. Another advantage is that this criterion may be used in both probabilistic and deterministic approaches, whereas the current approaches are mostly deterministic (with deterministic calculations used only for estimates of uncertainty ranges of input parameters).

  • 37. Fischer, M.
    et al.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bezlepkin, V. V.
    Hamazaki, R.
    Miassoedov, A.
    Core melt stabilization concepts for existing and future LWRs and associated R&D needs2015In: International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015, 2015, Vol. 9, p. 7578-7592Conference paper (Refereed)
    Abstract [en]

    In the event of a severe accident with core melting in a NPP the stabilization of the molten corium is an important mitigation issue, as it can avoid late containment failure caused by basemat penetration, overpressure, or severe damage of internal structures. The related failure modes may result in significant long-term radiological consequences and high related costs. Because of this, the licensing framework of several countries now includes the request to implement mitigative core melt stabilization measures. This does not only apply to new builds but also to existing LWR plants. The paper gives an overview of the ex-vessel core melt stabilization strategies developed during the last decades. These strategies are based on a variety of physical principles like: melt fragmentation in a deep water pool or during molten core concrete interaction with top-flooding, water injection from the bottom (COMET concept), and retention in an outside-cooled crucible structure. The provided overview covers the physical background and functional principles of these concepts, as well as their status of validation and, if applicable, the remaining open issues and R&D needs. For concepts based on melt retention inside a cooled crucible that reached sufficient maturity to be implemented in current Gen-III+ designs, like the VVER-1000/1200 and the EPR™, more detailed descriptions are provided, which include key aspects of the related technical realization. The paper is compiled using contributions from the main developers of the individual concepts.

  • 38. Fischer, Manfred
    et al.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bezlepkin, Vladimir V.
    Hamazaki, Ryoichi
    Miassoedov, Alexei
    Core Melt Stabilization Concepts for Existing and Future LWRs and Associated Research and Development Needs2016In: NUCLEAR TECHNOLOGY, ISSN 0029-5450, Vol. 196, no 3, p. 524-537Article in journal (Refereed)
    Abstract [en]

    In the event of a severe accident in a nuclear power plant with the core melting, the stabilization of the molten corium is an important mitigation issue, as it can avoid late containment failure caused by basemat penetration, overpressure, or severe damage to internal structures. The related failure modes may result in significant long-term radiological consequences and related high costs. Because of this, the licensing frameworks of several countries now include a requirement to implement mitigative core melt stabilization measures. This applies not only to new builds but also to existing light water reactors. The paper gives an overview of the ex-vessel core melt stabilization strategies developed during the last decades. These strategies are based on a variety of physical principles, like melt fragmentation in a deep water pool or during the molten core-concrete interaction with top flooding, water injection from the bottom (COMET), and retention in an outside-cooled crucible structure. This overview covers the physical background and functional principles of these concepts, as well as their validation status and, if applicable, the remaining open issues and research and development needs. For the concepts based on melt retention inside a cooled crucible that have reached sufficient maturity to be implemented in current Generation III+ designs, like the VVER-1000/1200 and the European Pressurized Water Reactor, more detailed descriptions are provided, which include key aspects of the related technical realization. The paper is compiled using contributions from the main developers of the individual concepts.

  • 39. Granovsky, V. S.
    et al.
    Khabensky, V. B.
    Krushinova, E. V.
    Vitol, S. A.
    Sulatsky, A. A.
    Almjashev, V. I.
    Bechta, Sevostian V.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Gusarov, V. V.
    Barrachin, M.
    Bottomle, P. D.
    Fischer, M.
    Piluso, P.
    Oxidation effect on steel corrosion and thermal loads during corium melt in-vessel retention2014In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 278, p. 310-316Article in journal (Refereed)
    Abstract [en]

    During a severe accident with core meltdown, the in-vessel molten core retention is challenged by the vessel steel ablation due to thermal and physicochemical interaction of melt with steel. In accidents with oxidizing atmosphere above the melt surface, a low melting point UO2+x-ZrO2-FeOy corium pool can form. In this case ablation of the RPV steel interacting with the molten corium is a corrosion process. Experiments carried out within the International Scientific and Technology Center's (ISTC) METCOR Project have shown that the corrosion rate can vary and depends on both surface temperature of the RPV steel and oxygen potential of the melt. If the oxygen potential is low, the corrosion rate is controlled by the solid phase diffusion of Fe ions in the corrosion layer. At high oxygen potential and steel surface layer temperature of 1050 degrees C and higher, the corrosion rate intensifies because of corrosion layer liquefaction and liquid phase diffusion of Fe ions. The paper analyzes conditions under which corrosion intensification occurs and can impact on in-vessel melt retention (IVR).

  • 40. Granovsky, V. S.
    et al.
    Sulatsky, A. A.
    Khabensky, V. B.
    Sulatskaya, M. B.
    Gusarov, V. V.
    Almyashev, V. I.
    Komlev, A. A.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kim, Y. S.
    Park, R. J.
    Kim, H. Y.
    Song, J. H.
    Modeling of melt retention in EU-APR1400 ex-vessel core catcher2012In: International Congress on Advances in Nuclear Power Plants 2012, ICAPP 2012, 2012, p. 1412-1421Conference paper (Refereed)
    Abstract [en]

    A core catcher is adopted in the EU-APR1400 reactor design for management and mitigation of severe accidents with reactor core melting. The core catcher concept incorporates a number of engineering solutions used in the catcher designs of European EPR and Russian WER-1000 reactors, such as thin-layer corium spreading for better cooling, retention of the melt in a water-cooled steel vessel, and use of sacrificial material (SM) to control the melt properties. SM is one of the key elements of the catcher design and its performance is critical for melt retention efficiency. This SM consists of oxide components, but the core catcher also includes sacrificial steel which reacts with the metal melt of the molten corium to reduce its temperature. The paper describes the required properties of SM. The melt retention capability of the core catcher can be confirmed by modeling the heat fluxes to the catcher vessel to show that it will not fail. The fulfillment of this requirement is demonstrated on the example of LBLOCA severe accident. Thermal and physicochemical interactions between the oxide and metal melts, interactions of the melts with SM, sacrificial steel and vessel, core catcher external cooling by water and release of non-condensable gases are modeled.

  • 41.
    Grishchenko, Dmitry
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Basso, Simone
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Sensitivity Study of Steam Explosion Characteristics to Uncertain Input Parameters Using TEXAS-V Code2014In: NUTHOS10, Paper-1293, Okinawa, Japan, 2014, Atomic Energy Society of Japan , 2014Conference paper (Refereed)
    Abstract [en]

    Release of core melt from failed reactor vessel into a pool of water is adopted in several existing designs of light water reactors (LWRs) as an element of severe accident mitigation strategy. Corium melt is expected to fragment, solidify and form a debris bed coolable by natural circulation. However, steam explosion can occur upon melt release threatening containment integrity and potentially leading to large early release of radioactive products to the environment. There are many factors and parameters that could be considered for prediction of the fuel-coolant interaction (FCI) energetics, but it is not clear which of them are the most influential and should be addressed in risk analysis. The goal of this work is to assess importance of different uncertain input parameters used in FCI code TEXAS-V for prediction of the steam explosion energetics. Both aleatory uncertainty in characteristics of melt release scenarios and water pool conditions, and epistemic uncertainty in modeling are considered. Ranges of the uncertain parameters are selected based on the available information about prototypic severe accident conditions in a reference design of a Nordic BWR. Sensitivity analysis with Morris method is implemented using coupled TEXAS-V and DAKOTA codes. In total 12 input parameters were studied and 2 melt release scenarios were considered. Each scenario is based on 60,000 of TEXAS-V runs. Sensitivity study identified the most influential input parameters, and those which have no statistically significant effect on the explosion energetics. Details of approach to robust usage of TEXAS-V input, statistical enveloping of TEXAS-V output and interpretation of the results are discussed in the paper. We also provide probability density function (PDF) of steam explosion impulse estimated using TEXAS-V for reference Nordic BWR. It can be used for assessment of the uncertainty ranges of steam explosion loads for given ranges of input parameters.

  • 42.
    Grishchenko, Dmitry
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Konovalenko, Alexander
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Karbojian, Aram
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinova, Valtyna
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Insight into steam explosion in stratified melt-coolant configuration2013In: 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, 2013Conference paper (Refereed)
    Abstract [en]

    Release of core melt from failed reactor vessel into a pool of water is adopted in several existing designs of light water reactors (LWRs) as an element of severe accident  mitigation  strategy.  When  vessel  breach  is  large  and  water  pool  is shallow,  released  corium  melt  can  reach  containment  floor  in  liquid  form  and spread under water creating a stratified configuration of melt covered by coolant. Steam  explosion  in  such  stratified  configuration  was  long  believed  as  of secondary importance for reactor safety because it was assumed that considerable mass of melt cannot be premixed with the coolant. In this work we revisit these assumptions  using  recent  experimental  observations  from  the  stratified  steam explosion tests  in  PULiMS  facility.  We  demonstrate  that  (i)  considerable  melt-coolant premixing layer can be formed in the stratified configuration with high temperature  melts,  (ii)  mechanism  responsible  for  the  premixing  is  apparently more  efficient  than  previously  assumed  Rayleigh-Taylor  or  Kelvin-Helmholtz instabilities.  We  also  provide  data  on  measured  and  estimated  impulses, energetics  of  steam  explosion,  and  resulting  thermal  to  mechanical  energy conversion ratios. 

  • 43.
    Gusarov, V.V.
    et al.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Almjashev, V.I.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Bechta, Sevostian
    A.P. Alexandrov Research Institute of Technology (NITI).
    Influence of the Temperature Difference at Immiscibility Liquids Interface on their Phase Instability2007In: Proceedings of the MASCA2 Seminar 2007, 2007Conference paper (Refereed)
  • 44.
    Gusarov, V.V.
    et al.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Almjashev, V.I.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Bechta, Sevostian
    Alexandrov Research Institute of Technologies (NITI).
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Method for producing ceramic materials incorporating ferric oxide, alumina, and silicon dioxide for nuclear-reactor molten core trap2002Patent (Other (popular science, discussion, etc.))
    Abstract [en]

    NOVELTY - Production of ceramic materials for nuclear reactor molten core trap involves preparation of charge by milling and mixing components, and producing molding powder from charge. This molding powder is molded and formed into briquettes, which are calcined in an air environment at 1300 - 1380 degreesC for 2-14 hr. The briquettes are crushed, the powder is milled and sized to fractions, and mixed with binder. The parts are molded and calcined at 1200-1300 degreesC for 4-14 hr. Silicon dioxide and part of the alumina are introduced into the charge as kaolin, whose content is higher than the desired content of silicon dioxide in material by 2.1-2.3 times. USE - In nuclear power engineering, for confining molten core in water-cooled tank reactors. ADVANTAGE - Enhanced reproducibility of physical and mechanical properties of molten core sacrificial ceramic materials.

  • 45.
    Gusarov, V.V.
    et al.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Almjashev, V.I.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Bechta, Sevostian
    Aleksandrov Research Institute of Technology.
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    Distribution of components between immiscible melts of a system under nonisothermal conditions2006In: Glass Physics and Chemistry, ISSN 1087-6596, E-ISSN 1608-313X, Vol. 32, no 6, p. 638-642Article in journal (Refereed)
    Abstract [en]

    The influence of the temperature difference at the interface on the redistribution of components between coexisting liquid phase is analyzed using the U-Zr-O system as an example. It is demonstrated that, in this system, there can arise new dynamic effects in boundary regions of the coexisting phases. These effects are of considerable practical importance, for example, for the prediction of the behavior of the system in severe accidents at nuclear power plants.

  • 46.
    Gusarov, V.V.
    et al.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Almjashev, V.I.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Bechta, Sevostian
    Aleksandrov Research Technological Institute.
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    Interaction of a material based on aluminum and iron oxides with a metal melt2007In: Russian journal of applied chemistry, ISSN 1070-4272, E-ISSN 1608-3296, Vol. 80, no 4, p. 528-535Article in journal (Refereed)
    Abstract [en]

    Interaction of an oxide material with a melt of metals in the combustion mode was studied experimentally. The conditions in which processes of this kind can occur without any increase in the temperatureof the environment are possible were analyzed.

  • 47.
    Gusarov, V.V.
    et al.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Almjashev, V.I.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Bechta, Sevostian
    A.P. Alexandrov Research Institute of Technology (NITI).
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    Physicochemical Modeling and Analysis of the Interaction between a Core Melt of the Nuclear Reactor and a Sacrificial Material2005In: Glass Physics and Chemistry, ISSN 1087-6596, E-ISSN 1608-313X, Vol. 31, no 1, p. 53-66Article in journal (Refereed)
    Abstract [en]

    -The problems regarding the design of a new class of materials, namely, sacrificial materials for usein devices intended for localization of a core melt of the nuclear reactor, are considered. Criteria are proposedfor the proper choice of the chemical composition of a sacrificial material, as well as of the composition andmicrostructure of its constituents. The possible alternatives are outlined and analyzed. The results of designinga variant of the composition of sacrificial materials are presented. The experimental data on the interaction ofan oxide sacrificial material with simulators of the metal and oxide phases of the core melt are discussed. A newtype of combustion of materials, namely, the liquid-phase combustion, is revealed. It is demonstrated that thematerial designed can be used in systems intended for localization of a core melt of the nuclear reactor.

  • 48.
    Gusarov, V.V.
    et al.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Almjashev, V.I.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Bechta, Sevostian
    Aleksandrov Research Institute of Technology.
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    Physicochemical simulation of the combustion of materials with the total endothermal effect2007In: Glass Physics and Chemistry, ISSN 1087-6596, E-ISSN 1608-313X, Vol. 33, no 5, p. 492-497Article in journal (Refereed)
    Abstract [en]

    A new type of combustion, namely, the combustion of materials without heating of the environment, is described, and the conditions under which this process can occur are analyzed. It is demonstrated that the possibility of occurring the process under consideration depends substantially on the microstructure of the material. The characteristics of the material for which the liquid-phase combustion takes place without an increase in the temperature of the melt are determined using the interaction of a material based on iron and aluminum oxides with the Fe-Zr-O melt as an example.

  • 49.
    Gusarov, V.V.
    et al.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Almjashev, V.I.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Bechta, Sevostian
    A.P. Alexandrov Research Institute of Technology (NITI).
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    Новый класс функциональных материалов для устройства локализации расплава активной зоны ядерного реактора2005In: Russian chemical journal, Vol. 49, no 4, p. 42-53Article in journal (Refereed)
  • 50.
    Gusarov, V.V.
    et al.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Almjashev, V.I.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Stoliarova, V.L.
    Bechta, Sevostian
    Alexandrov Research Institute of Technologies (NITI).
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Oxidic material for nuclear reactor core melt catcher2001Patent (Other (popular science, discussion, etc.))
    Abstract [en]

    NOVELTY - Oxide material of molten core catcher incorporating Al2O3,SiO2, also contains Fe2O3 and/or Fe3O4, and target dopant of one or more oxides of following group: SrO,CeO2,BaO,Y2O3,La2O3. Proportion of ingredients is, in mass percent: Fe2O3 and/or Fe3O4, 46-80; Al2O3, 16-50; SiO2, 1-4; target dopant, 3-15. The material has content of radioactive strontium and cerium isotopes in gas phase above molten core that is reduced by 2-7 times. USE - For nuclear reactor molten core catcher. ADVANTAGE - Reduced radioactivity of molten core.

12 1 - 50 of 79
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