PROJECT PROPOSAL	#????

I. Summary Project Information

1. Project Title and Taxonomy

Full title:	Thermo-hydraulics of U-Zr-O molten pool under oxidising conditions in multi-scale approach (crucible - bundle - reactor scales).
Short title:	THOMAS (Thermo-Hydraulics of Oxidising Melt in Severe () Accidents)
Technology area:	Nuclear Reactor Safety
Category of technology development:	Basic Research
Key words:	Severe accidents, Nuclear Power Plants, corium melt, thermohydraulics, oxidation

2. Project Manager

Name:	Mikhail S. Veshchunov
Title:	Dr., Professor	Position:	Head of laboratory
Street address:	B.Tulskaya Street, 52
City:	Moscow	Region:
ZIP:	115191	Country:	Russia
Tel.:	+7 (495) 955-2218	Fax:	+7 (495) 958-0040
E-mail:	vms@ibrae.ac.ru

3.Participating Institutions

3.1. Leading Institution: IBRAE

Full name:	Nuclear Safety Institute of Russian Academy of Science
Street address:	B.Tulskaya Street, 52
City:	Moscow	Region:
ZIP:	115191	Country:	Russia
Signature Authority:	Leonid A. Bolshov
Title:	Dr., Professor	Position:	Director
Tel.:	+7 (495) 952-2421	Fax:	+7 (495) 958-0040
E-mail:	bolshov@ibrae.ac.ru
Governmental Agency:	Russian Academy of Science

4. Non-CIS Collaborators

4.1. Collaborator 1: FZK

Institution:	Forschungszentrum Karlsruhe GmbH
Street address:	P.O. Box 3640
City:	Karlsruhe	Region/State:
ZIP:	D-76021	Country:	Germany
Person:	Walter Tromm
Title:	Doctor	Position:	Deputy Head Nuclear Safety Research Programme
Tel.:	07247/825510	Fax:	07247/825508
E-mail:	walter.tromm@psf.fzk.de
Person:	Alexei Miassoedov
Title:	Doctor	Position:	Group leader
Tel.:	+49 (0)7247/822553	Fax:	+49 (0)7247/822095
E-mail:	alexei.missoedov@iket.fzk.de
Person:
Title:	Position:
Tel.:	Fax:
E-mail:

4.2. Collaborator 2: …..

5. Project Duration

36 months.

6. Project Location and Equipment

Institution	Location, Facilities and Equipment
Leading Institution IBRAE	WorkStations, Computers

7. Total Person-Months of Project Effort

Total number of participants	20	17
Number of weapon scientists and engineers	10	9
Total project effort (person*days)	9518	5868
Total project effort of weapon scientists and engineers (person*days)	4960	2174

8. Financial Information

8.1. Estimated Project Costs

Estimated total cost of the project (US $)	310’000
Including:
Payments to Individual Participants	270’000
Equipment	-
Materials	-
Bank Fees	-
Other Direct Costs	5’000
Travel	30’000
Overheads	5’000

8.2. Funding Sources

Estimated total cost of the project (US $)	310’000
Financial Sources:
Requested from the ISTC	310’000
Other financial source 1	-
Other financial source 2	-
Non-Financial Sources:
Non-financial source 1	-
Non-financial source 2	-

9. Summary of the project

The physical phenomena involved in severe accidents at Nuclear Power Plants are extremely complex. They generally demand the development of specific research. The research in this area aims to further understanding of these physical phenomena and to reduce the uncertainties surrounding their quantification. In this field, there is no way to conduct experiments on full scale and to reproduce all the possible situations. Therefore the ultimate goal is to develop models, which are validated on the basis of scaled experiments and which then can be applied to the reactor cases. These models, grouped together in computer codes, should allow predicting severe accident progression.

Non-destructive and destructive post-test examinations of bundles in various tests showed formation of molten pools of different scales at various stages of core degradation. Small local pools were observed at upper elevations in bundles in the early stage of core degradation in CORA and QUENCH tests. Results of the Phebus FP tests confirmed that a significant part of the fuel bundle was liquefied and that the amount of fuel damage was close to TMI-2 with an extended molten pool located in a central zone of the bundle underneath a cavity. In the late stage of a severe accident, melt can relocate into the lower head of the reactor pressure vessel and form a large molten pool interacting with cooled walls. The details of in-vessel melt pool behaviour are preconditions for possible in-vessel melt retention considerations. The in-vessel melt retention aspects and the improvement of the predictability of the thermal loading are a matter of especially high interest and priority for BWR and reactors with low power density (e.g. VVER-440), but also for large PWR and VVER reactors.

Oxidation kinetics of Zr-containing melts can be significantly higher in comparison with that of solid materials, therefore, it strongly determines high heat generation and hydrogen source term during severe accidents, as unambiguously shown, e.g., in the QUENCH tests. On the other hand, the oxidation kinetics strongly depends on thermal hydraulic behaviour of oxidised melt. Therefore, there is a critical need to investigate the in-vessel molten pool behaviour under oxidation conditions and to predict the changes corium will undergo, including the main phenomena, which can govern these changes: natural convection in the molten pool; corium oxidation including H2 production; metal/oxide stratification in the molten pool; dissolution of reactor vessel steel at sub-melting point temperatures.

The objective of the project is to perform the modelling work on a mechanistic description of U-Zr-O molten pool behaviour in oxidising conditions, and to extend the thermal hydraulic consideration of oxidised melt from small scale (crucible tests) up to a large scale (reactor pressure vessel), including an intermediate scale corresponding to molten pools in the bundle tests.

Scope of the work:

Further development and improvement of the physico-chemical model for the U-Zr-O melt oxidation developed at IBRAE within various International projects on the base of crucible tests, and later extended to a larger scale tests CORA, QUENCH and Phebus FP;

Further development and improvement of the unified thermal hydraulic technique (CONV code) for simulation of multiphase processes in complex domains of convectively stirred melt developed at IBRAE within various International projects on the base of experiments with a heat generating fluids (MASCA, RASPLAV and LIVE);

Tight coupling of the two advanced tools: the physico-chemical (molten pool oxidation) model and the thermo-hydraulic code CONV, for realistic mechanistic description of U-Zr-O molten pool behaviour in oxidising test conditions;

Extension of the thermal hydraulic consideration of oxidised melt from small scale (crucible tests) up to a large scale (reactor pressure vessel), including an intermediate scale corresponding to molten pools in the bundle tests; this includes improved interpretation of the bundle Phebus FP tests observations of corium melt oxidation, as well as transposition of thermal hydraulic consideration from the MASCA, RASPLAV and LIVE experiments to the reactor case.

State of the art. In the main system codes, which are presently used for the analysis of reactor core degradation under severe accident conditions, the models dealing with melt oxidation are oversimplified and are not able to give the adequate description of the main phenomena, which can govern the changes corium will undergo in oxidising conditions, especially in application to the real reactor scale.

Impact of the proposed project on the progress in the field of severe accident modelling. The proposed modelling work should significantly improve the understanding of the molten pool oxidation processes as well as enabling better use of the currently existing and future experimental data in this area. This, in turn, will improve the modelling of reactor core molten materials behaviour under severe accident conditions.

Competence of the project teams. The IBRAE specialists have gained considerable experience in the above-specified area of research. During the last 15 years original investigations have been performed at IBRAE concerning melt relocation, physico-chemical interactions (such as molten Zr oxidation, ZrO2 and UO2 dissolution by the melt), molten pool convection and heat exchange.

The proposed project work will use previously obtained results: theoretical models and numerical modules for mechanistic description of melt oxidation and UO2 dissolution applied for analysis of small-scale (crucible) and bundle tests; numerical CONV code (developed at IBRAE) describing turbulent hydrodynamics and heat exchange of the molten pool.

Expected results and their application. It is assumed that advanced physical models and numerical modules describing the in-vessel molten pool behaviour under oxidation conditions and predicting the changes corium will undergo, will be developed and implemented in the thermal-hydraulic code CONV. The modified code will be will be verified against available experimental data and applied to the analysis of the physico-chemical behaviour of molten pool in the reactor pressure vessel. The developed and verified models will be further used for benchmarking and improvement of simplified models of various system codes such as ICARE/CATHARE, MELCOR, ASTEC or Russian severe accident code SOCRAT.

PROJECT PROPOSAL	#

II. Detailed Project Information

1. Introduction and Overview

Severe accidents at Nuclear Power Plants can cause significant damage to reactor fuel resulting in more or less complete core meltdown [1]. Such accidents are highly unlikely in light of the preventive measures implemented by operators. However, they are the focus of considerable research, because the release of radioactive products into the environment would have serious consequences. This research also reflects a commitment to the defense-in-depth approach.

The physical phenomena involved in severe accidents are extremely complex [2,3]. They generally demand the development of specific research. The research in this area aims to further understanding of these physical phenomena and to reduce the uncertainties surrounding their quantification. In this field, there is no way to conduct experiments on full scale and to reproduce all the possible situations. Therefore the ultimate goal is to develop models, which are validated on the basis of scaled experiments and which then can be applied to the reactor cases. These models, grouped together in computer codes, should allow predicting severe accident progression.

Non-destructive and destructive post-test examinations of bundles in various tests showed formation of molten pools of different scales at various stages of core degradation. Small local pools were observed at upper elevations in bundles in the early stage of core degradation in CORA [4,5] and QUENCH tests [6,7]. Results of the Phebus FP tests [8,9] confirmed that a significant part of the fuel bundle was liquefied and that the amount of fuel damage was close to TMI-2 [10] with an extended molten pool located in a central zone of the bundle underneath a cavity. In the late stage of a severe accident, melt can relocate into the lower head of the reactor pressure vessel and form a large molten pool interacting with cooled walls. The details of in-vessel melt pool behaviour are preconditions for possible in-vessel melt retention considerations. The in-vessel melt retention aspects and the improvement of the predictability of the thermal loading are a matter of especially high interest and priority for BWR and reactors with low power density (e.g. VVER-440), but also for large PWR and VVER reactors [3].

Oxidation kinetics of Zr-containing melts can be significantly higher in comparison with that of solid materials, therefore, it strongly determines high heat generation and hydrogen source term during severe accidents, as unambiguously shown, e.g., in the QUENCH tests [6,7]. On the other hand, the oxidation kinetics strongly depends on thermal hydraulic behaviour of oxidised melt. Therefore, there is a critical need to investigate the in-vessel molten pool behaviour under oxidation conditions and to predict the changes corium will undergo, including the main phenomena, which can govern these changes: natural convection in the molten pool; corium oxidation including H2 production; metal/oxide stratification in the molten pool; dissolution of reactor vessel steel at sub-melting point temperatures.

Oxidation kinetics of Zr melt was initially investigated at FZK in small-scale crucible tests [11]. Qualitative results of those studies allowed better understanding and interpretation of corium oxidation in CORA and QUENCH tests. Within the ISTC Project #2936 a great effort was undertaken for the development of numerical physico-chemical model for the U-Zr-O melt oxidation. Being developed on the base of the crucible tests, the new model was extended later to a larger scale and allowed quantitative interpretation of observations of CORA, QUENCH and Phebus tests ("molten pool oxidation"). The model development was initiated in the frames of the COLOSS Project (5th FP) [11-13] whereas its implementation in the SVECHA code and application to the bundle tests [14, 15] were strongly supported by SARNET NoE (6th FP) in collaboration with JRC/IE and FZK.

The main deficiency of the SVECHA approach was simplified thermal hydraulic description of the convectively stirred melt. Similarly to modelling of crucible tests, in this approach the main thermal hydraulic characteristics of the convectively stirred melt were taken into consideration in the form of simple correlations for the mass transfer coefficients at the oxide/melt interface. However, such important predictions of the new model for the melt interacting (at its open surface) with steam as formation of a thick oxide crust on cold walls (owing to oxygen transport from the open surface to the walls through the convectively stirred melt under non-equilibrium temperature conditions) which can essentially suppress corium coolability, require more detailed and self-consistent consideration of mass- and heat-exchange processes in the melt, especially in the prototypic (large-scale) geometry of the reactor pressure vessel. These important phenomena can be adequately modelled only by tight coupling of the physico-chemical melt oxidation model with advanced thermal hydraulic codes.

In parallel, the 3D unified thermal hydraulic technique for simulation of multiphase processes in complex domains of convectively stirred melt were strongly advanced in the CONV code within the same ISTC Project #2936. Initially developed in the framework of RASPLAV and MASCA [] Projects, the code was further improved and validated in the ISTC Project using experimental data, including experiments with a heat generating fluids like LIVE, COPO, SIMECO [16, 17].

Numerical approach and computing module for simulation of phase changes in pure materials and binary mixtures were developed and tested against a melting of pure gallium problem [18]. On this base, a new micro-segregation module for modelling of mushy zone growth is under development.

For multi-dimensional simulation of various thermal-hydraulic processes in NPP in a wide range of parameters (Ra < 1016 and Re 103-104) the 3D unified CFD numerical technique was developed. Results of numerical testing in the area of turbulent flows allow a possibility of DNS (Direct Numerical Simulation) and LES (Large Eddy Simulation) modelling approach to the large-scale thermal hydraulic problems [19].

The proposed Project aims at the tight coupling of the two advanced tools developed within the previous Project #2936: the SVECHA physico-chemical (molten pool oxidation) model and the thermo-hydraulic code CONV. This will allow a realistic mechanistic description of U-Zr-O molten pool behaviour in oxidising conditions and will extend the thermal hydraulic consideration of oxidised melt from small scale (crucible tests) up to a large scale (reactor pressure vessel), including an intermediate scale corresponding to molten pools in the bundle tests. Moreover, improved interpretation of Phebus FP tests observations of corium melt oxidation, as well as transposition of thermal hydraulic consideration from the experiments (e.g. MASCA, RASPLAV) to reactor case, are foreseen.

2. Expected Results and their Application

The category of technology development of the research proposed is basic research.

Within the framework of the proposed project it is assumed that advanced physical models and numerical modules describing the in-vessel molten pool behaviour under oxidation conditions and predicting the changes corium will undergo, will be developed and implemented in the thermal-hydraulic code CONV. The modified code will be will be verified against available experimental data (e.g. Phebus FP tests (IRSN, Cadarache)) and applied to the analysis of the physico-chemical behaviour of molten pool in the reactor pressure vessel. The solution of this task will be complementary to the LIVE (Late In-Vessel phase Experiments) project in FZK, Karlsruhe and thus will provide analytical support of the LIVE test.

The proposed work is complementary to the SARNET-II Project, the obtained results and new models can be used also in this Project. New crucible tests on corium melt oxidation planned at the PLINIUS Corium Platform (CEA, France), will be used for verification of the melt oxidation model.

The developed and verified models will be further used for benchmarking and improvement of simplified models of various system codes such as ICARE/CATHARE, MELCOR, ASTEC or Russian severe accident code SOCRAT.

The new knowledge about reactor core degradation expected from the project will make it possible to improve the accident management measures and risk assessment.

Expected long term consequences of successful project completion: progress in the fuel technology and its operation and in the design of the next generation reactors of improved safety.

In the future the project team can participate in the implementation of the developed models in the existing system codes.

3. Meeting ISTC Goals and Objectives

The proposed project provides weapons scientists opportunities to redirect their talents to peaceful activities, promotes integration of Russian scientists into the international scientific community, supports basic research and technology development for peaceful purposes in the field of nuclear safety.

4. Scope of Activities

Task 1. Development and Improvement of the Physico-Chemical Model for the U-Zr-O Melt Oxidation on the Base of New Crucible Tests (1-6 QUATERS)

Task description and main milestones	Participating Institutions
1.1. The existing model for U-Zr-O melt oxidation will be updated to the 2-d case corresponding to a more realistic geometry of molten pool in crucible and bundle tests. On this base, pre-test calculations for the planned new crucible tests on corium melt oxidation at the PLINIUS Corium Platform (CEA, France) will be carried out. As a result, a matrix for the test series will be prepared in cooperation with collaborators from CEA and FZK (Germany). The work will be carried out within quarters 1-2, and completed by release of the report. 1.2. Analysis and post-test calculations of the new PLINIUS tests will be carried out. On this base, further modification and improvement of the corium melt oxidation model will be carried out. The work will be carried out within quarters 3- 4, and completed by release of the report. 1.3. For consideration of molten corium physico-chemical interactions with reactor walls, a model for U-Zr-O melt interaction with stainless steel under non-equilibrium conditions with temperature gradient in the cooled walls, will be developed on the base of available experimental data and coupled with the melt oxidation model. The work will be carried out within quarters 5-6, and completed by release of the report.	IBRAE
Description of deliverables
1	Quarterly Reports
2	Annual Reports

Task 2. Development and Improvement of the Unified Thermal Hydraulic Technique (CONV Code) for Simulation of Multiphase Processes in Complex Domains of Convectively Stirred Melt (1-6 QUATERS)

Task description and main milestones	Participating Institutions
2.1. Modernization of the existing code CONV: including into code of a new procedure for solving elliptical equations with boundary conditions of Neumann for the pressure correction and of a newly developed computing scheme of the highest order of accuracy for solving of the Navier-Stokes equations system. The work will be carried out within 1-2 quarters and completed by release of the report. 2.2. Conducting of numerical experiments for the choice of optimum turbulence model (algebraic type) using results of both convection in a cavity with the walls with different temperatures and convection of a heat-generating fluid, and model implementation in the CONV code. The work will be carried out within 3-4 quarters and completed by release of the report. 2.3. Improvement of the turbulence model: large eddy simulation approach will be combined with algebraic turbulence model. Selection of a set of commutative filters for a LES approach of a turbulence modeling and its implementation in the CONV code. Adaptation of the code to the design of the LIVE facility. The work will be carried out within 5-6 quarters and completed by release of the report.	IBRAE
Description of deliverables
1	Quarterly Reports
2	Annual Reports

Task 3. Tight Coupling of the Two Advanced Tools: the Physico-Chemical (Molten Pool Oxidation) Model and the Thermo-Hydraulic Code CONV, for Realistic Mechanistic Description of U-Zr-O Molten Pool Behaviour in Oxidising Test Conditions (7-10 QUATERS)

Task description and main milestones	Participating Institutions
3.1. The modified melt oxidation model will be upgraded to the geometry of molten pool confined by the walls of reactor pressure vessel. An interface of the melt oxidation numerical module for coupling with the thermal hydraulic code CONV will be elaborated. The work will be carried out within quarters 7-8 and completed by release of the report. 3.2. Preparatory work with the code CONV 2D sources for implementation of the physico-chemical melt oxidation model, development of an interface (input files) with the melt oxidation model. The work will be carried out within quarters 7-8 and completed by release of the report. 3.3. Implementation of the physico-chemical melt oxidation model in the code CONV 2D for modeling of thermal hydraulic behavior of U-Zr-O melt under oxidising conditions for small and medium scale experiments. The work will be carried out within quarters 9-10 and completed by release of the report.	IBRAE
Description of deliverables
1	Quarterly Reports
2	Annual Reports

Task 4. Extension of the Thermal Hydraulic Consideration of Oxidised Melt from Small Scale (Crucible Tests) up to a Large Scale (Reactor Pressure Vessel), Including an Intermediate Scale Corresponding to Molten Pools in the Bundle Tests (11-12 QUATERS)

Task description and main milestones	Participating Institutions
4.1. Adaptation of the physicochemical melt oxidation model to modeling of thermal hydraulic behaviour of U-Zr-O melt under oxidising conditions for separate elements of the large-scale geometry (reactor pressure vessel). The work will be carried out within quarters 9-10 and completed by release of the final report. 4.2. Validation and verification of the melt oxidation model and modified version of the CONV code on accessible experimental data and benchmark tests (e.g. Phebus FP tests). The work will be carried out within quarters 11-12 and comes to an end by release of the final report.	IBRAE
Description of deliverables
1	Quarterly Reports
2	Annual Reports

5. Role of Foreign Collaborators

CEA (Cadarche, France) and FZK (Karlsruhe, Germany) will perform new tests on corium melt oxidation in steam at the PLINIUS Corium Platform (CEA), which will be used for verification of the melt oxidation model.

FZK will also provide experimental data from LIVE (Late In-Vessel phase Experiments) project concerning 3-D heat flux distribution to the lower head and possible resulting crust formation, gap formation between the vessel wall and the melt crust, the influence of the melt relocation mode to the lower head on melt pool formation and heat flux distribution.

IRSN/DPAM (Cadarache, France) will collaborate in benchmarking of the simplified models against the advanced models produced by the project, in order to have a benefit for the ICARE/CATHARE and ASTEC codes involved in the modelling/ validation effort in SARNET.

The three institutes will participate in information exchange, cross-checks of results obtained in the course of project implementation and in technical monitoring of project activities performed by ISTC staff; provide comments to the technical reports, submitted by project participants to the ISTC.

6. Technical Approach and Methodology

6.1. Physico-Chemical Model for Corium Melt Oxidation

Oxidation kinetics of Zr melt was initially investigated in FZK small-scale crucible tests and theoretically analysed by IBRAE in [11]. Qualitative results of those studies allowed better understanding and interpretation of corium oxidation in QUENCH tests [6,7]. The model explained the formation of ceramic precipitates induced by the temperature difference between solid and liquid materials, and predicted continuous oxidation/precipitation process after attainment of the saturation state of the melt. Furthermore, the model predicted a linear or close to linear time law for the rate of the ZrO2 ceramic phase (oxide layer + precipitates) growth during Zr melt oxidation that corresponds to a much faster kinetics of Zr oxidation and hydrogen generation in comparison with the standard (parabolic) rate. Numerical calculations of the model allowed a qualitative interpretation of the vigorous melt oxidation and ZrO2 phase precipitation observed in several of the QUENCH tests. Additional analysis of the model showed that dissolution of the oxide layer by the melt under thermal non-equilibrium conditions can proceed after attainment of the melt saturation. This model prediction was successfully confirmed by the FZK tests specially designed for investigation of such dissolution behaviour at a late stage of crucible-melt interactions [11].

For quantitative analysis of melt behaviour in the bundle tests, melt oxidation should be considered self-consistently with the UO2 pellets dissolution. A new model for these processes was developed at IBRAE based on the new RIAR (Dimitrovgrad, Russia) tests on simultaneous dissolution of UO2 and ZrO2 by molten Zry, performed in the frame of the COLOSS Project in collaboration with JRC/IE [12]. However, analysis of the RIAR tests showed that the temperature difference between the crucible wall and melt, initially suspected to be large, was in fact relatively small. For this reason, a simplified isothermal model was numerically realised in that Project.

The melt oxidation model was further generalised by IBRAE in collaboration with JRC/IE (Petten) [14] for description of the ternary U-Zr-O melt oxidation. It was shown that the main qualitative conclusion on the enhanced kinetics of melt oxidation and hydrogen generation was also valid for the ternary system. Numerical calculations of the model allowed qualitative interpretation of the vigorous melt oxidation and (U,Zr)O2-x phase precipitation observed in the CORA tests [4,5].

The subsequent 2-d model [15] for non-equilibrium U-Zr-O melt oxidation was developed at IBRAE in collaboration with JRC/IE (Petten) on the base of the previously developed 1-d models [14] and extended to consideration of the general case of simultaneous UO2 dissolution and melt oxidation accompanied with growth of the peripheral ceramic layer (crust) and bulk ceramic precipitates. The model was extended to the general case considering temperature differences between the melt and two solid phases (pellet and peripheral oxide layer surrounding melt) in the course of their non-equilibrium interactions. This allowed self-consistent treatment of UO2 dissolution and melt oxidation accompanied with growth of the peripheral ceramic layer (crust) and bulk ceramic precipitates. The model was validated against isothermal crucible tests [11] where precipitation of ceramic phase during molten Zr interactions with ceramic crucible walls was observed along with oxide growth on non-oxidised (free) surface of the melt. In order to analyse melt oxidation under transient temperature conditions of the Phebus FP tests [8,9], the new physico-chemical model was tightly coupled with the heat exchange model and applied to interpretation of the post-test microstructure observations of the molten pool in the FP bundles.

In the previous ISTC Project #2936 [] a preliminary study of proposed tests for oxidation in steam of corium melt confined in ceramic crucible with cooled walls (i.e. in geometry similar to the reactor pressure vessel) was carried out. An important model predictions for the melt interacting (at its open surface) with steam was derived concerning formation of a thick oxide crust on cold walls (owing to oxygen transport from the open surface to the walls through the convectively stirred melt under non-equilibrium temperature conditions) which can essentially suppress corium coolability. It was outlined that analysis of melt oxidation in the prototypic (large-scale) geometry of the reactor pressure vessel required more advanced and self-consistent consideration of mass- and heat-exchange processes in the convectively stirred melt with high Raleigh numbers. These complicated phenomena can be adequately modelled only by tight coupling of the physico-chemical melt oxidation model with the advanced thermal hydraulic code CONV; this important task was recommended for the new ISTC Project.

6.2. Thermal Hydraulic Code System

During the last 15 years, 3D unified thermal-hydraulic technique for simulation of two-phase processes in complex domains is being developed in IBRAE [16-19], [21-30].

The developed technique includes:

methods, algorithms and software for automatic generation of Cartesian grids with local refinement near to solid-state borders,

methods and algorithms for solving of heat and mass transfer incompressible problems for research of 3D thermal hydraulic phenomena,

approaches for turbulence modelling.

For grid construction an automatic technology using CAD systems for designing of calculation domains has been developed. A generation of structured orthogonal/Cartesian grids with a local refinement near boundaries is incorporated into specially developed program, which is supplied by a user-friendly interface and can be used on parallel computers.

To solve computational fluid dynamics problems in the domain of arbitrary shapes including the variable properties of materials the new effective finite-volume numerical algorithm has been developed, which is applied to the heat and fluid flow equations (i.e. Navier-Stokes equations with energy equation) in the primitive variables formulation. A computing technique is based on the developed algorithms with small scheme diffusion, for which the discrete approximations are constructed using finite-volume methods and fully staggered grids.

A decomposition principle on physical processes is used to construct operator-splitting schemes.

In the momentum equation the operators are split into two parts. The first part is associated with the velocity transport by convection/diffusion, whereas the second part deals with pressure gradient. The Numerical implementation of the constructed scheme was performed as the predictor-corrector procedure. As a result, introduction of the pressure correction in the momentum equations allowed it to be transformed to the well-known Poisson equation and velocity correction equation.

The operators of energy equation are decomposed into two parts associated with the enthalpy and temperature. Fully implicit scheme (backward time difference) is used for the unsteady energy equation.

For the solution of convection problem the regularized nonlinear monotonic operator-splitting scheme was developed.

The developed computing technique includes the fictitious domain method (FDM) to deal with irregular computing areas and a modified Preconditioned Richardson Method with Fast Fourier Transformation (FFT) as a preconditioner for the solution of the pressure equation. This combination gives more than a ten times acceleration in comparison with standard Preconditioned Conjugate Gradient Methods (PCGM) of incompressible CFD problems.

The developed scheme provides the second order in space and the first order in time. The proposed algorithm is stable at sufficiently large steps of time-integration.

For the modelling of 3D turbulent single-phase flows algebraic turbulence model is used. Use of this model needs choice of parameters which will be defined from the set of verification problems.

For the modeling of 3D turbulent multiphase flows direct numerical simulation (DNS) approach will be applied also. To implement model of convective flows simultaneously with heat exchange to cooled boundary and crust formation at high Rayleigh numbers with the help LES (large eddy simulation) approach is planned also.

The previously developed in the frameworks of the ISTC project #2936 the three-dimensional CONV code for simulation of three-dimensional flows in the Boussinesq approximation on Cartesian grids with a local refinement near domain with singularities of flows was supplemented model for calculation of flows with a variable density without Boussinesq approximation.

On the base of the carried out numerical researches and adaptation of previously developed software, more perfect version of the code permitting calculation of three-dimensional flows with variable properties, appropriate to real corium properties or prototypes was obtained.

Numerical predictions of flows in a boundary layer conjugate with cooled boundaries were carried out. The results were supported by the theoretical analysis and have shown a high degree of a prediction of a considered class of flows by the CONV code. Similar investigations were carried out for examination of heat generating fluid behavior depending on boundary conditions on the upper boundary. The relevant estimates of heat transfer also demonstrate a high prediction power of the CONV code.

Model for numerical simulation of melt flows simultaneously with heat transfer in the reactor case and crust formation at high Rayleigh numbers (Ra 1016) was developed and implemented in the code. Matrix of validation for developed models was created, including as results of the theoretical analysis and experimental data (BALI, COPO and SIMECO).

On the basis of the modified CONV version, simulation of molten-salt tests obtained at the RASPLAV facility was carried out. The obtained results demonstrate that the developed approach predicts correctly the absolute values of the crust thickness.

The preliminary numerical simulation of circular heat used in the experimental facility LIVE was carried out by means of the CONV code which was adapted to the experimental conditions.

The elaborated technique has a high degree of efficiency and at the moment allows CFD calculations to be carried out on a grid with a total number of 107 nodes using a PC with P4-2.0GHz processor and 1.5Gb-RAM. For simulation of the single-phase flows fast actions of 10-5 second per node and time step are possible.

Algorithms, methods and software, elaborated on the basis of the above computational methods and schemes have already been validated on a wide class of test problems, and were successfully used for the numerical support of 3D experiments conducted in the framework of RASPLAV and MASCA projects (sponsored by OECD).

7. Technical Schedule

Quarter 1	Quarter 2	Quarter 3	Quarter 4	Quarter 5	Quarter 6	Quarter 7	Quarter 8	Quarter 9	Quarter 10	Quarter 11	Quarter 12	Person*days
Task 1 Development and Improvement of the Physico-Chemical Model for the U-Zr-O Melt Oxidation	Report
Sub task 1.1 Update of the 2-d model and pre-test calculations of the new crucible tests	Technical report
Sub task 1.2 Post-test analysis and model improvement	Technical report
Sub task 1.3 Development of a model for oxidised corium melt interactions with stainless steel	Technical report
Person*days
Task 2 Development and Improvement of the Unified Thermal Hydraulic Technique (CONV Code)	Report
Sub task 2.1 Modernization of an existing code CONV.	Technical report
Sub task 2.2 Conducting of numerical experiments for the choice of optimum model of a turbulence (algebraic type).	Technical report
Sub task 2.3 Improvement of the turbulence model: coordination of algebraic model with LES. Selection of a set of commutative filters for LES approach.	Technical report
Person*days
Task 3 Tight Coupling of the Two Advanced Tools	Report
Sub task 3.1 Development of interface of melt oxidation numerical module for coupling with CONV code	Technical report
Sub task 3.2 Preparatory work with the code sources for implementation of the oxidation model in CONV 2D.	Technical report
Sub task 3.3 Implementation of the oxidation model in CONV 2D code.	Technical report
Person*days
Task 4 Extension of the Thermal Hydraulic Consideration from Small Scale (Crucible Tests) up to Large Scale (Reactor Pressure Vessel)	Final report
Sub task 4.1 Adaptation of the oxidation model to separate elements of large-scale geometry (reactor pressure vessel).	Technical report
Sub task 4.2 Testing of melt oxidation model and modified CONV code version against available experimental data.	Technical report
Person*days
TOTAL

8. Personnel Commitments

8.2. Managerial responsibilities

Project manager Prof. Dr. M.S. Veshchunov Laboratory for Theory of Condensed Matter Head: Prof. Dr. M.S. Veshchunov Laboratory for Computational Heat and Hydrodynamics Head: Dr. V.V. Chudanov

The organisational structure of the project and the managerial responsibilities.

9. Financial Information

9.1. Estimated Project Costs (US $)

Estimated total cost of the project	310’000
Leading Institution	310’000

9.1.1. Payments to Individual Participants (US $)

Institution	Category I	Category II	Supporting personnel	Total
Leading Institution	270’000
Subtotal:	270’000

9.1.2. Equipment

Institution	Equipment description	Cost (US $)
Leading Institution	-
Subtotal:	-

9.1.3. Materials

Institution	Materials description	Cost (US $)
Leading Institution	-
Subtotal:	-

9.1.4. Other Direct Costs

Institution	Direct costs description	Cost (US $)
Leading Institution	Technological energy, communications	5’000
Subtotal:	5’000

9.1.5. Travel costs (US $)

Institution	CIS travel	International travel	Total
Leading Institution	-	30’000	30’000
Subtotals:	30’000

9.1.6. Overhead (US $)

Institution	Amount
Leading Institution	5’000
Subtotal:	5’000

9.2. Funding Sources

Estimated total cost of the project (US $)	310’000

9.2.1. Financial Sources

Financial Source	Written confirmation (Y/N)	Amount (US $)
Requested from the ISTC	310’000

9.2.2. Non-Financial Sources

Source	Short description of contribution	Written confirmation (Y/N)	Estimated amount (US $)
-

9.2.3. Submitted for Funding to Program beside the ISTC

No funding beside the ISTC is expected.

10. Intellectual Property Statement

The rights for intellectual property that are generated during the course of the project will be regulated by the laws of the Russian Federation and by the procedures, which have been developed by the ISTC.

The general conditions on Intellectual Property Rights as described in the Model Project Agreement will be observed.

11. Monitoring and Auditing Statement

In accordance with Article VIII of the ISTC Agreement, project recipients will give to the Center and to each Party which wholly or partly finances a project the right of access to carry out on-site monitoring and audit of all activities of the project. Project agreements will specify the portions of facilities, equipment, documentation, information, data systems, materials, supplies, personnel, and services which will concern the project and therefore will be made accessible for monitoring and audit. Project recipients shall have the right to protect those portions of facilities that are not related to the project.

12. Supporting Information

Supporting letters from…

References

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Chudanov V.V., Aksenova A.E., Pervichko V.A., 3D unified CFD approach to modeling of bubble phenomena. Proc. The 11th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-11) Popes’ Palace Conference Center, Avignon, France, October 2-6, 2005. CDROM paper 047. 2005.

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V.V. Chudanov, A.G. Churbanov, A.G. Popkov, V.F. Strizhov, P.N. Vabishchevich, Numerical simulation of heat and mass transfer processes at severe accidents of nuclear reactors. In Programming and Mathematical Techniques in Physics (Eds. Yu.Yu.Lobanov and E.P.Zhidkov), pp.186-188, World Scientific, Singapure, 1994.

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