Alexandrov Research Institute of Technology NITI

Saint Petersburg, Russia

Alexandrov Research Institute of Technology NITI

Saint Petersburg, Russia
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Bottomley D.,Itu Institute For Transurane | Stuckert J.,KIT Campus Nord | Hofmann P.,KIT Campus Nord | Tocheny L.,ISTC Krasnoproletarskaya 32 34 | And 21 more authors.
Nuclear Engineering and Design | Year: 2012

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. © 2012 Elsevier B.V. All rights reserved.

Granovsky V.S.,Alexandrov Research Institute of Technology NITI | Khabensky V.B.,Alexandrov Research Institute of Technology NITI | Krushinov E.V.,Alexandrov Research Institute of Technology NITI | Vitol S.A.,Alexandrov Research Institute of Technology NITI | And 8 more authors.
Nuclear Engineering and Design | Year: 2014

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-ZrO 2-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 °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). © 2014 The Authors.

Bechta S.V.,Alexandrov Research Institute of Technology NITI | Krushinov E.V.,Alexandrov Research Institute of Technology NITI | Vitol S.A.,Alexandrov Research Institute of Technology NITI | Khabensky V.B.,Alexandrov Research Institute of Technology NITI | And 13 more authors.
Nuclear Engineering and Design | Year: 2010

Qualitative and quantitative determination of the release of low-volatile fission products and core materials from molten oxidic corium was investigated in the EVAN project under the auspices of ISTC. The experiments 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 oxidation kinetics, critical influence of melt surface temperature and oxidation index on the fission product release rate, aerosol particle composition and size distribution. The relevance of measured high release of Sr from the molten pool for the reactor application is highlighted. Comparisons of the experimental data with those from the COLIMA CA-U3 test and the VERCORS tests, as well as with predictions from IVTANTHERMO and GEMINI/NUCLEA codes are made. Recommendations for further investigations are proposed following the major observations and discussions. © 2010 Elsevier B.V. All rights reserved.

Sulatsky A.A.,Alexandrov Research Institute of Technology NITI | Smirnov S.A.,D.V. Efremov Scientific Research Institute of Eelectrophysical Apparatus | Granovsky V.S.,Alexandrov Research Institute of Technology NITI | Khabensky V.B.,Alexandrov Research Institute of Technology NITI | And 10 more authors.
Nuclear Engineering and Design | Year: 2013

Experimental, theoretical and numerical studies of oxidation kinetics of an open surface corium pool have been reported. The experiments have been carried out within OECD MASCA program and ISTC METCOR, METCOR-P and EVAN projects. It has been shown that the melt oxidation is controlled by an oxidant supply to the melt free surface from the atmosphere, not by the reducer supply from the melt. The project experiments have not detected any input of the zirconium oxidation kinetics into the process chemistry. The completed analysis puts forward a simple analytical model, which gives an explanation of the main features of melt oxidation process. The numerical modeling results are in good agreement with experimental data and theoretical considerations. © 2013 Published by Elsevier B.V.

Granovsky V.S.,Alexandrov Research Institute of Technology NITI | Sulatsky A.A.,Alexandrov Research Institute of Technology NITI | Khabensky V.B.,Alexandrov Research Institute of Technology NITI | Sulatskaya M.B.,Alexandrov Research Institute of Technology NITI | And 8 more authors.
International Congress on Advances in Nuclear Power Plants 2012, ICAPP 2012 | Year: 2012

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.

Bogdanov R.V.,Saint Petersburg State University | Kuznetsov R.A.,Saint Petersburg State University | Epimahov V.N.,Alexandrov Research Institute of Technology NITI | Titov A.V.,Saint Petersburg State University | Prudnikov E.E.,Saint Petersburg State University
Recent Patents on Engineering | Year: 2013

The paper presents new polyphase ceramic waste forms (matrices) of aluminosilicate-phosphate type synthesized from natural bauxites and apatite ore tailings. This ceramic material is named "geoceramics" by the authors. The optimum composition of the waste matrix is selected (Cs2O, P2O5, SiO2, and Al2O3) and a cost-saving method for synthesis of matrices capable to accommodate up to 12 wt. % of cesium and 6 wt. % of strontium isotopes is developed. Phases which immobilize cesium isotopes are identified. It is shown that waste forms which are close in stoichiometric composition to pollucite (CsAlSi2O6) have the best resistance to water. The rate of cesium leaching (R) from these waste forms is 2·10-6 g/cm2day in the kinetic region. Heat treatment and cooling of the matrix material does not affect the immobilization performance of the proposed waste forms. In some cases, a positive effect is achieved by using the sol-gel method which can reduce the grain sizes, resulting in a decrease in the leach rate of cesium to 1·10-6 g/cm2day. The leach rate of strontium is below the detection limit of atomic absorption spectrophotometers (0.3 · 10-6 g/cm2day). © 2013 Bentham Science Publishers.

Gricay A.S.,Alexandrov Research Institute of Technology NITI | Migrov Y.A.,Alexandrov Research Institute of Technology NITI
Thermal Engineering (English translation of Teploenergetika) | Year: 2015

The article deals with matters concerned with the problem of determining the statistical characteristics of variable parameters (the variation range and distribution law) in analyzing the uncertainty and sensitivity of calculation results to uncertainty in input data. A comparative analysis of modern approaches to uncertainty in input data is presented. The need to develop an alternative method for estimating the uncertainty of model parameters used in thermal–hydraulic computer codes, in particular, in the closing correlations of the loop thermal hydraulics block, is shown. Such a method shall feature the minimal degree of subjectivism and must be based on objective quantitative assessment criteria. The method includes three sequential stages: selecting experimental data satisfying the specified criteria, identifying the key closing correlation using a sensitivity analysis, and carrying out case calculations followed by statistical processing of the results. By using the method, one can estimate the uncertainty range of a variable parameter and establish its distribution law in the above mentioned range provided that the experimental information is sufficiently representative. Practical application of the method is demonstrated taking as an example the problem of estimating the uncertainty of a parameter appearing in the model describing transition to post-burnout heat transfer that is used in the thermal–hydraulic computer code KORSAR. The performed study revealed the need to narrow the previously established uncertainty range of this parameter and to replace the uniform distribution law in the above mentioned range by the Gaussian distribution law. The proposed method can be applied to different thermal–hydraulic computer codes. In some cases, application of the method can make it possible to achieve a smaller degree of conservatism in the expert estimates of uncertainties pertinent to the model parameters used in computer codes. © 2015, Maik Nauka-Interperiodica Publishing, all rights reserved.

Yudov Yu.V.,Alexandrov Research Institute of Technology NITI | Danilov I.G.,Alexandrov Research Institute of Technology NITI | Chepilko S.S.,Alexandrov Research Institute of Technology NITI
Kerntechnik | Year: 2015

The Russian KORSAR/GP (hereinafter KORSAR) computer code was developed by a joint team from Alexandrov NITI and OKB "Gidropress" for VVER safety analysis and certified by the Rostechnadzor of Russia in 2009. The code functionality is based on a ID two-fluid model for calculation of two-phase flows. A 3D CFD module in the KORSAR computer code is being developed by Alexandrov NITI for representing 3D effects in the downcomer and lower plenum during asymmetrical loop operation. The CFD module uses Cartesian grid method with cut cell approach. The paper presents a numerical algorithm for coupling ID and 3D thermal-hydraulic modules in the KORSAR code. To ensure mass and energy balances at the interface, convective fluxes for ID module are calculated as a sum of fluxes at corresponding boundary faces of 3D module. Mass balance condition is used to couple pressures in the modules. The combined pressure field is calculated by the multigrid method. The performance efficiency of the algorithm for coupling ID and 3D modules was demonstrated by solving the benchmark problem of mixing cold and hot flows in a T- junction. © Carl Hanser Verlag, München.

Elshin A.,Alexandrov Research Institute of Technology NITI
Lecture Notes in Computer Science (including subseries Lecture Notes in Artificial Intelligence and Lecture Notes in Bioinformatics) | Year: 2015

This paper describes an application of the surface harmonics method to derivation of few-group finite difference equations for neutron flux distribution in a 3D triangular-lattice reactor model. The Boltzmann neutron transport equation is used as the original equation. Few-group finite difference equations are derived, which describe the neutron importance distribution (the multiplication factor in the homogeneous eigenvalue problem) in the reactor core. The derived finite difference equations remain adjoint to each other like the original equation of neutron transport and its adjoint equation. Non-diffusion approximations apply to calculation of a whole reactor core if we increase the number of trial functions for describing the neutron flux distribution in each cell and the size of the matrices of the few-group coefficients for finite difference equations. © Springer International Publishing Switzerland 2015.

Elshin A.,Alexandrov Research Institute of Technology NITI
Physics of Reactors 2016, PHYSOR 2016: Unifying Theory and Experiments in the 21st Century | Year: 2016

This paper develops an algorithm for deriving a set of finite difference equations (on an unstructured mesh) to describe neutron fields in heterogeneous reactor cores. The proposed approach is based on the surface harmonics method (SHM). The SHM uses the neutron transport equation in integral differential form to construct finite different equations, the step of deriving the diffusion equation in differential form being omitted. Finite difference equations are derived for describing both neutron flux distribution and neutron importance distribution. Using test problems as an example we demonstrate necessity and possibility of SHM passing to more exact approximations.

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