24th Symposium of AER on VVER Reactor Physics and Reactor Safety

NUCLEAR FUEL FOR NPPs WITH VVER: A STATUS AND PERSPECTIVE DESIGN
Dolgov A.B., Kukushkin Y.A., Ugryumov A.V.
JSC “TVEL”, Kashirskoe shosse 49, 115409 Moscow, Russia Kukushkin@tvel.ru
ABSTRACT
Carbon-free power engineering with its most developed part, nuclear power, is becoming more and more promising in the world power engineering. One of the most important issues of power engineering remain – improving safety level and stability in NPP operation and its economic efficiency, limited natural uranium resources, accumulation of spent nuclear fuel and its transporting and storage, as well as radwastes disposal problem.
VVER core reliability at normal operation, anticipated operational occurrences and design basic accidents is defined by nuclear fuel and its components characteristics as well as matching of Safety Analysis Report of the reactor unit and NPP as a whole to normative requirements and regulations in force in nuclear power field.
TVEL Fuel Company TVEL of Rosatom provides nuclear fuel to 76 power reactors in 15 countries of the world; namely VVER-440 and VVER-1000.
TVEL Fuel Company of Rosatom consists of the following blocks: – R&D;
- gas centrifuge manufacturing;
- nuclear fuel conversion and enrichment;
- nuclear fuel fabrication.
TVEL’s main goals in nuclear fuel fabrication are:
- meeting Customers’ requirements regarding operational characteristics and improving nuclear fuel technical and economic performance;
- providing necessary safety level during nuclear fuel operation and fabrication;
- improving competitiveness as well as marketing.
The said goals are achieved by solving the following tasks:
1. Development of fuel assemblies’ design which meets the requirements of reliable, safe and economically effective operation.
2. Development of new and improvement of existing fuel assemblies and design materials.
3. Improvement of fuel assemblies’ and core components’ manufacturing technologies.
4. Decreasing conservatism at VVER core substantiation, developing of calculation programs.
Recently as a result of the R & D and technological works performed by the leading plants of Russia’s nuclear power industry new types of nuclear fuel have been developed, introduced and are being operated at VVER NPP for:
- VVER-440 – second generation fuel with increased uranium enrichment per U-235
- VVER-1000 – ТВСА-PLUS and ТВС-2M and their modifications.
Technical and economic characteristics of VVER nuclear fuel correspond to the level of world manufactures of nuclear fuel for PWR reactors.
Further work to develop and improve designs of VVER fuel rods and fuel assemblies, design and fuel materials is being done.

DIFFERENT FUEL OPTIONS FOR THE 15-MONTH CYCLES OF PAKS NPP UNITS
Imre Nemes, István Pós, Paks NPP Ltd, Sándor Patai Szabó , TS Enercon Ltd., Hungary
ABSTRACT
4 Units of Paks NPP are planned to operate in 15 month cycles. The number of fresh fuel planned for this cycles is 102/unit/cycle. In this work we tried to improve fuel demand of 15m cycles, the goal was to establish 90 assembly/cycle strategies. For this purpose different developed assembly structure was investigated. Appropriate enrichment is determined, CCS libraries was generated to different versions and equilibrium cycles are arranged. Fresh fuel cost and benefits are calculated for each version.

VVER-1000 FUEL CYCLES: CURRENT SITUATION AND PERSPECTIVE
Volkov E.S., Kosourov E.K., Pavlov V.I., Pavlovichev A.M., Spirkin E.I, Shcherenko Anastasia I.
NRC “Kurchatov Institute”, Moscow, Russia

Over the last years works were underway on increasing of FA uranium capacity and the thermal power for operating VVER-1000 reactor units and on development of new projects with use of VVER-1000 technologies. In the report current state of the fuel cycle (core configurations, general neutronics characteristics and other) is described, examples of fuel cycles that are used and planned to use on operating reactor units and NPPs which are under construction are given. Problems of fuel cycles choice for maintenance the possibility of units to work in maneuverable mode are considered.
Implementation features (in terms of neutronics characteristics) of the 18-month fuel cycle under reactor operation on 104 % of the nominal power level are given. The results of optimization of burnable poison distribution within the FA are discussed in respect to this operating mode.
In the conclusion the prospect of used and proposed engineering solutions for improvement of the VVER-1000 fuel cycle is assessed.

NEW FUEL CYCLE STRATEGIES FOR VVER-440
Gagarinski A.A., Ivanova E.G., Lizorkin M.P., Proselkov V.N. NRC “Kurchatov Institute”, Moscow, Russia
ABSTRACT
Wide implementation of second-generation VVER-440 (V-213) reactors is characteristic for the contemporary phase of development of medium water-water reactors, which are in operation at NPPs such as Loviisa (Finland), Paks (Hungary), Dukovany (Czech Republic), Mochovce and Bohunice (Slovakia), Rovno (Ukraine), Kola (Russia) and Metsamor (Armenia). Two new VVER-440 units (3 and 4) will be launched at the Mochovce NPP in 20152016. So the improvement of fuel and fuel cycles for VVER-440 is an important task today.
Contemporary fuel cycle strategies for VVER-440 are different compared to those used in 20002010.
The basic features of the contemporary VVER-440 fueling strategy are:

• capability to operate at an uprated thermal power level with fuel load lifetime of at least 340 effective days;
• fuel burnup limited by 68 MWday/kgU (average per fuel rod);
• optimized fuel enrichment profile with respective reduction of the initial fuel enrichment;
• lower conservatism of fuel cycle validation. To meet these tasks, it is necessary to:
• increase the share of fuel to be replaced in the course of refueling, in order to allow an uprated thermal power operation;
• shift from a six-year fuel cycle to a four- or five-year one.
  This paper presents relevant results for fuel cycles with second-generation assemblies, which were developed for Slovak NPP units to enable their operation at an uprated power level of 109%.
  These results were obtained for VVER-440 fuel cycles based on RK2+ assemblies with higher U content and RK3 assemblies, assuming the reactor to operate at uprated thermal power levels of 107% (1471 MW), 109% (1498 MW) and 112% (1540 MW).
  Advantages of the fuel cycle based on RK2+ assemblies with U-235 enrichment higher than 5% are also analyzed.

INFORMATION ABOUT AER WG A ON IMPROVEMENT, EXTENSION AND VALIDATION OF PARAMETRIZED FEW-GROUP LIBRARIES FOR VVER 440 AND VVER 1000
Mikoláš Pavel,
ŠKODA JS a.s., Orlík 266, 316 06 Plzeň, Czech Republic
Petr Darilek, VUJE Inc., Slovakia
ABSTRACT
Joint AER Working Group A on „Improvement, extension and validation of parameterized few-group libraries for VVER-440 and VVER-1000“ and AER Working Group B on „Core design“ twenty second meeting was hosted by VÚJE a.s. in Modra-Harmónia, Slovakia, during the period of April 14 to 16, 2014.
In total 15 participants from 10 member organizations took part in the meeting and 11 papers were presented.
Objectives of the meeting of WG A are: Issues connected with spectral calculations and few- groups libraries preparation, their accuracy and validation.
Presentations were devoted to some aspects of transport and diffusion calculations, and especially to the benchmark dealing with VVER-440 core periphery power tilt.
M. Gren presented paper “Two Group Constants Generation for ANDREA Code Using SCALE System”, Gy. Hegyi (co-authors E. Temesváry, Cs. Maráczy and G. Hordósy) spoke about “Some Additional Comments on the Results of the VVER440 Benchmark“, P. Mikoláš added “Remarks to the Full Core Benchmark“. P. Mikoláš (co-author J. Švarný) also spoke about “VVER In-Core Self-Powered Neutron Detector Calculation Benchmark – Partial Solution”, then I. Pós (co-authors T. Parkó and S. P. Szabó) presented paper “Study of Pin- wise Models“, finally J. Bajgl supplied “Remarks to Subcritical Multiplication Calculations Using Different MOBYDICK Code Options”.
During the meeting were discussed questions connected with the benchmark solution and, among others, future activities, which are also shortly described in the end of the paper.

AER WORKING GROUP B ACTIVITIES IN 2014
Petr Darilek
VUJE Inc., Okruzna 5, SK 918 64 Trnava, Slovakia
petr.darilek@vuje.sk
Review of AER Working Group B Meeting in Modra – Harmónia, Slovakia is given.

DEVELOPMENT OF CODES AND «KASKAD» COMPLEX
Lizorkin M.P.
NRC «Kurchatov institute», Moscow, Russia
ABSTRACT
The brief description of development history of engineering computer codes for the VVER reactors calculations been developed in VVER physics department of NRC «Kurchatov institute», since the end of 50th years of the last century till now, will be given in the report.
The modern status of the basic codes of "Kurchatov institute» for the VVER fuel loadings design and safety analysis (TVS-M, KASKAD-BIPR-PERMAK, BIPR8, PERMAK-3Д, SC1, TOPRA, the ATHLET/BIPR-VVER, TIGR-1, etc.) from the various points of view will be characterised:

• Mission of codes and a solved problems;
• Interaction of codes;
• Interesting features of calculational models;
• Degree of verification and reliability of calculations; * Accuracy of calculation of target parameters;
• Lacks and problems, perspective directions of codes development.
  Some specific problems which are shown during designing of fuel cycles, perfection of fuel and technologies will be commented. They demand decision within the complex of programs and, accordingly, demand the further development of a complex or working out and inclusion in a complex of new programs. Such problems concern:
  *VVER calculations at using the spectral regulation of criticality (instead of boric regulation);
• Definition pin-by-pin power distribution in transients (including fast processes), etc. In the report the information on creation of the KASKAD user Club, the Club format,
  and also conditions of participation in current work and single actions of Club will be given.

DEVELOPMENT OF METHODICAL APPROACHES IN VVER CALCULATIONS
Bolobov P.A., Lazarenko A.P., Tomilov M.Ju. NRC “Kurchatov Institute”, Moscow, Russia
ABSTRACT
The report provides a brief description of the current state of the KASKAD software package intended for VVERs neutronic calculations. The issues related to various sources of calculation error are considered including:
– the spectral code errors;
– uncertainties associated with some features of the calculation model used in the preparation of few-group constants (for instance, the use of constant values of the state parameters when calculating the FA depletion);
– errors of few-group constant approximation scheme;
– the coarse-mesh calculation uncertainties (for instance, neglecting the effect of 10B depletion in the soluble boron during the fuel loading operation).
The report gives a special attention to various aspects of improving methodological approaches used in calculations.

ACCOUNTING FOR IMPACT OF FA GAP DEVIATIONS IN VVER-1000 TO ENGINEERING MARGIN FACTORS
Shishkov L.K, Gorodkov S.S., Mikailov E.F., Sukhino-Homenko E.A, Sumarokova A.S. NRC «Kurchatov Institute», Russia, Moscow
ABSTRACT
Jacket-free fuel assemblies of VVER-1000 and their western analogues change the shape in the process of operation. Typically, we note longitudinal bending. FA bendings threaten hazard, such as, first of all, reduced speed of control rod drop, however, alongside with induced bending there changes (also increases) moderator amount in the gaps between fuel assemblies, which results in additional jump of power release.
Nowadays, FA design has been supplied with amendments that ensure design speeds of CR drop. Nevertheless, FA shape changes and this effect shall be taken into account in calculation of power density. Gap values are of random character and their impact shall be accounted using engineering margin factors. In case of VVER-1000, application of margin factor leads to fuel cost increase in the cost of electrical power by (3 – 5) %, while the half of the cost increment is stipulated by changed gaps between fuel assemblies.
The paper discusses the mechanism of accounting gap influence on power density in tvels, gives numerical values of power coefficients of sensitivity and gap values calculated by selected computational models, and discusses the interference of gap influence and accounting for feedbacks.

ASSESSMENT OF THE UNCERTAINTIES OF MULTICELL CALCULATIONS BY THE OECD NEA UAM PWR PIN CELL BURNUP BENCHMARK
István Panka, András Keresztúri
Centre for Energy Research, Hungarian Academy of Sciences Reactor Analysis Department
H-1525 Budapest 114, P.O. Box 49, Hungary istvan.panka@energia.mta.hu
ABSTRACT
It is well known that best-estimate analyses have a lot of benefits in the field of nuclear energy, although the corresponding calculations have to be supplemented by the assessment of the uncertainties. For this purpose an OECD NEA benchmark was set up. The objective of the OECD NEA Uncertainty Analysis in Best-Estimate Modeling (UAM) LWR benchmark is to determine the uncertainties of the coupled reactor physics/thermal hydraulics LWR calculations at all stages.
At the beginning, the uncertainties of the neutronic calculations at zero burnup were investigated in the frame of the UAM benchmark. That time was devoted to the elaboration of the appropriate methods quantifying the uncertainties of calculations. Later on the benchmark has been extended for the burned out case, as well. In this paper the focus is on the determination of the burnup dependent uncertainties of MULTICELL spectral calculations for a PWR pin cell defined in the benchmark.
The uncertainties of the MULTICELL calculations are discussed using Monte-Carlo type statistical analyses taking into account the energy dependent covariance matrices of the cross sections. Additionally, the impact of the uncertainties of the fission yields is also considered. The target quantities are the burnup dependent uncertainties of the infinite multiplication factor, the two-group cross sections and reaction rates and the number densities of some isotopes up to the burnup of 60 MWd/kgU. In the paper the burnup dependent tendencies of the corresponding uncertainties and their reasons are analyzed.

INVESTIGATION OF EFFECT PRODUCED BY FLUID TRANSVERSE FLOWS BETWEEN VVER FUEL ASSEMBLIES TO CONSERVATISM OF THERMAL-HYDRAULIC ANALYSIS IN A CONSISTENT SOLUTION OF NEUTRONIC AND THERMAL-HYDRAULIC TASKS
Shumskiy B.E., Kotsarev A.V., Lizorkin M.P. NRC “Kurchatov Institute”, Moscow, Russia
ABSTRACT
The report describes the agreed three-dimensional neutronic and thermal-hydraulic models of calculation of the VVER reactor core for software complexes NOSTRA and TRETON and complex ATHLET/BIPR-VVER. Using these models, studies the influence of transverse flow of fluid between the FA on the thermo physical parameters (temperature of coolant, of fuel, of fuel rod cladding, departure from nucleate boiling) used in the analysis of thermal reliability and safety of operation of fuel rod. The results of the calculation of thermal-hydraulic parameters obtained in the framework of the in-channel and three- dimensional models of the VVER-1000 core are compared.
When the agreed three-dimensional thermo-hydraulic and neutron-physical calculations of processes in the VVER core with fuel assemblies without covers the effect of feedbacks can rebuild the field of energy release adversely in relation to the characteristics of thermal reliability. This leads to the fact that the results of the thermo physical parameters calculation obtained using the three-dimensional model of heat hydraulics are more conservative in relation to the results of in-channel models. The effect of the coolant flow rate value through the core to the difference between the results of the in-channel and three-dimensional calculation also are studied.

3-D SIMULATION OF VVER CORE POWER DENSITY BASED ON SAPFIR_95&RC_VVER ALGORITHMS
Artemov V.G., Kuznetsov A.N., Shemaev Yu.P. FSUE «Alexandrov NITI», Sosnovy Bor, Russia
ABSTRACT
SAPFIR_95&RC_VVER is certified by Federal Environmental, Industrial and Nuclear Supervision Service of Russia for calculation of VVER reactor neutronic characteristics. The calculation of subchannel power density is based on the superposition of micro-distribution of neutron flux in the fuel rod assembly and the macro-distribution in the complete reactor core (referred to as superposition method). The micro-distributions are calculated by the SAPFIR_95 program on the basis of the solution of the transport equation in an infinite lattice of fuel rod assembly using the first-collision probability method. Nonuniformity of the neutron flux macro-distributions across the core is calculated by the RC_VVER program in the diffusive approximation.
In order to increase the calculation accuracy of individual rod characteristics and to obtain an error estimation of the calculations, the SAPFIR_95&RC_VVER package includes the possibility of solving the equation of neutron diffusion on hexagonal finite-difference grid corresponding to the fuel rod lattice (referred to as direct method).
The direct method of calculation is more complex and cumbersome, but it gives the detailed description of the core structure and side reflector. This method uses the solution of the equation of fine-grid diffusion corresponding to the real lattice pitch, leading to longer time required for calculation of one reactor state. Another and more important limitation of the direct method is a systematic accuracy of diffusive approximation in simulating the burnup of burnable poison rods and gadolinium-containing fuel rods.
Therefore, a combined calculation method of subchannel power density is developed and implemented in the SAPFIR_95&RC_VVER package. In this combined method, the superposition method calculates the burnup distributions which are used as input data for power density micro-field calculation. The superposition method also calculates the distributions of fuel temperature and coolant density which are also input data for the calculation of the subchannel power density by the direct method.
The paper presents the scheme of subchannel power density calculation by the combined calculation method implemented in the SAPFIR_95&RC_VVER. Results of comparative assessment are given which demonstrate the advantages of the combined method in comparison with the direct calculation and the superposition calculation method.

CALCULATION OF NEUTRON AND GAMMA-RAY DOSE RATES BY THE USE OF THE MONTE CARLO CODE MCU
Oleynik D.S.
NRC “Kurchatov institute”, Moscow, Russia,
ABSTRACT
Neutron and gamma-ray dose rate calculations are important part of the process of substantiation of radiation safety. Generally these rates are required in a point or surface at some distance from spent fuel.
The MCU code has been updated for that purposes, that is a point and ring detectors estimation and also surface flux tally have been included. Verification and validation of the code has been performed for VVER spent fuel storages and shipping casks using 38 measurements and 204 computational tasks. Uncertainty of the calculations has been formulated in the result of V&V.
Updated version of the MCU code provides neutron and gamma-ray dose rate calculations for substantiation of radiation safety of VVER spent fuel.

JUSTIFICATION OF PROPER DEGREE OF CONSERVATISM IN CALCULATION OF ENGINEERING MARGIN FACTORS FOR DNBR AS APPLIED TO VVER-1000 CORES CONTAINING FUEL ASSEMBLIES OF VARIOUS TYPES
Oleksyuk D.A., Pinegin A.A., Ryzhov A.A. NRC “Kurchatov Institute” Moscow, Russia
ABSTRACT
In the paper is represented the procedures of determining the engineering safety factors for DNBR, for the linear thermal load from the surface of the fuel element and for coolant heating based on complex statistical model neutron-physical and thermal-hydraulic processes in FA and thermomechanical deformation of FA. The proposed statistical model allows to eliminate a number of conservative assumptions that used in obtaining of engineering safety factors according to traditional methods and taking into account the correlations between the specified processes.
On the basis of the statistical model FA using FORTRAN-90 it was created STAT-TVS code, which includes SC-1 code as a block of thermal-hydraulic calculations. Statistical analysis of STAT-TVS code is performed by GRS method, which allows to refuse from hypothesis about normal distribution of the random parameters, which is important when taking into account the thermomechanical deformations of FA and test the robustness of results to the choice of statistical hypotheses.
Obtained by STAT-TVS code statistical estimates for DNBR, as a rule, is 15 % higher than the corresponding values obtained by traditional techniques.
In the paper is represented the results of sensitivity analysis of engineering safety factors for DNBR to change of power distribution in FA and used statistical hypotheses.
We consider the features of using engineering safety factors applied to non-stationary processes associated with the decrease of the power in FA and the reduction of coolant flow rate through the fuel assemblies. It is presented the first results of numerical researching.

METHODOLOGY OF 3D NEUTRON-PHYSICAL CALCULATION OF NUCLEAR REACTORS OF AXIAL-SYMMETRY WITH FINITE STEP ALONG THE AXIS
Poveschenko T.S., Laletin N.I., Sultanov N.V. NRC “Kurchatov Institute”, Moscow, Russia
ABSTRACT
In this paper methodology for 3-D calculation analysis of nuclear reactor cell with axial symmetry and finite mesh step is described. This methodology is based on the method of axial leakage calculation that has been developed for nuclear reactor with closed lattice like VVER. Trial functions used at full core level are defined. Analytical solutions of two-group diffusion equation with “symmetrical ” and ”cross” boundary conditions are given. Two-group cell characteristics that are matrixes of the boundary values of these solutions have been calculated. None-diagonal elements of this matrix are just the deductions caused by curvature of neutron flux. These values have been analyzed in dependence on the step mesh along Z (axis) value.
These results show that if mesh step along Z-axis is enough small (in comparison with neutron free path) then diagonal elements are closed to initial diffusion parameters of the cell and none-diagonal elements can be neglected. But in contrary if step mesh is more than diffusion length of heterogeneous medium then these deductions are essential and they should be taken into account at full core level.
This methodology is used in full core level with Surface Value System with finite step mesh along Z-axis in lattice code SVL.

METHODOLOGY OF CALCULATING THE AXIAL LEAKAGE FOR NUCLEAR REACTORS OF AXIAL SYMMETRY
Poveschenko T.S., Laletin N.I.
NRC “Kurchatov Institute”, Moscow, Russia
ABSTRACT
The problem of reactivity effect of coolant voiding is fundamental for theory and calculation analysis methods of nuclear reactors because it conditions reactor safety. Calculation analysis of axial leakage in nuclear reactor is very important part of this study. Especially it seems to be significant for VVER-type nuclear reactor because pitch size of the hexagonal lattice is small in comparison with neutron free path. So if we are going to model coolant voiding cell under accident conditions we have to do careful analysis of physical approximation of mathematical model. In this paper approach to solution of 3-D neutron transport equation is described. Direction collision probabilities are defined as a calculation tool to estimate axial diffusion coefficient Dz. Important aspects of boundary condition formulation on nuclear rector cell surface for the problem of axial leakage are investigated
Obtained results show that if influence of 2-D neutron flux spectrum into current along Z-axis is taken into account then axial leakage is changed in reflector area for VVER cell. Also the results show effect of increasing of axial diffusion coefficient when the closed lattice contains “void” gap. This effect vanishes if approximation of cylindrical cell is used and boundary conditions are formulated as isotropic reflection at the surface of equivalent cylindrical cell.

AVOIDING A TWICE ACCOUNTING OF MACRO-COMPONENT IN THE UNCERTAINTY OF ESTIMATED ENGINEERING MARGIN FACTORS FOR PARAMETERS OF VVER POWER DENSITY
Dementiev V.G., Gorodkov S.S., Mikailov E.F., Shishkov L.K., Sumarokova A.S., Sukhino-Homenko E.A.
NRC «Kurchatov Institute», Russia, Moscow
ABSTRACT
The analysis proves that traditional approaches to assessing engineering margin factors includes undue conservatism. In particular, some calculation uncertainties can be twice taken into account. For instance, MCU code evaluation of effect of changed gap between VVER- 1000 fuel assemblies to rod power also takes into account the changes in power macro-field induced by gap behaviour. Power macro-field contributes to calculated-and-measured deviation of calculated methodological uncertainty. Also, it is possible to point out that sometimes it happens that the contribution of technological deviations in the process of fuel fabrication to power macro-field is accounted in mechanical uncertainties though such contribution was already included in methodological uncertainties as these uncertainties are determined in the process of measurements at actual FAs.

Because of the difficulties with experimental validation of power distribution predicted by
macro-code on the pin-by-pin level we decided to prepare a calculation benchmark named
‘FULL-CORE’ VVER-440. This benchmark was a two-dimensional (2D) calculation
benchmark based on the VVER-440 reactor core cold state geometry with taking into account
the geometry of explicit radial reflector. The main task of this benchmark was to test the pinby-pin power distribution in fuel assemblies predicted by macro-codes that are used for
neutron-physics calculations especially for VVER-440 reactors.
The proposal of this benchmark was presented at the 21st Symposium of AER in 2011 ⁷.
The reference solution has been calculated by MCNP code using the Monte Carlo method and
the results have been published in the AER community. The results of reference calculation
were presented at the 22nd Symposium of AER in 2012 ⁸.
Comparison of the available macro-codes results of this calculation benchmark was presented
at the 23rd Symposium of AER in 2013 ⁹ and also in journal Kerntechnik 4/2014 ¹⁰.
In this contribution we define the ‘FULL-CORE’ VVER-440 benchmark extension, in which
CAs of working group 6 (7 assemblies in whole core) are fully inserted. Results requested by
resolvers are similar (identical) with the original benchmark.
The preliminary reference of the Monte Carlo calculation results are also provided.

VERIFICATION OF SAPFIR_95&RC_VVER PACKAGE FOR CORE NEUTRONIC CALCULATIONS IN EXTENDED RANGE OF PERMISSIBLE VALUES OF VVER PARAMETERS
Artemov V.G., Artemova L.M., Zinatullin R.E., Ivanov A.S., Kuznetsov A.N., Piskarev A.V., Shemaev Yu. P.,
FSUE «Alexandrov NITI», Sosnovy Bor, Russia
Kurakin K.Yu, Tihomirov A.N., Fateyev M.V. OKB «. Gidropress», Podolsk, Russia
ABSTRACT
The SAPFIR_95&RC_VVER package is designed for calculation of VVER neutronic characteristics, critical fuel assemblies, and computational polygrids of fuel storage pools.
The SAPFIR_95&RC_VVER package was certified by the Nuclear and Radiation Safety Center in 2005. It is now used in OKB “Gidropress” for both safety analysis of fuel loaded in operating reactors and design of advanced reactor cores.
In view of required increase in power of the operating VVERs and their longer operation period between refuelings, the design of the VVER fuel elements and assemblies was modified. An additional verification of the SAPFIR_95&RC_VVER package has been carried out to demonstrate that it is applicable to a safety analysis of cores with modified fuel taking into account changes in permissible values of fuel assembly and core parameters.
The additional verification has shown that neutronic calculation errors for the cores with modified fuel assemblies agree with errors given in the SAPFIR_95&RC_VVER Certificate 2005. Also, the verification has established lower calculation errors for critical boron concentration, efficiency of control rods, and worth of shutdown rods than those given in Certificate 2005.
The paper presents results of verification calculations on key neutronic parameters. The verification used experimental data on a series of fuel batches representing sequential introduction of new type fuel assemblies in the operating power units of the Balakovo NPP and the Rostov NPP. The neutronic calculation error for coolant boiling conditions is verified against experiments performed in the operating ВК-50 (BWR type) reactor. The calculation accuracy of power in individual fuel rods is verified against results of benchmark Monte- Carlo calculations.

APPLYING FULL SETS OF MULTIGROUP CELL CHARACTERISTICS OBTAINED BY MCU CODE IN FINITE-DIFFERENCE CALCULATIONS OF VVER NEUTRON FIELD
Gorodkov S.S., Kalugin M.A.
NRC “Kurchatov institute”, Moscow, Russia
ABSTRACT
Until recently it was possible to obtain with MonteCarlo approach only cell average crossections, which can be used either in finite difference calculations or as a benchmark values for some approximate spectral codes. Now MonteCarlo based MCU code can tally certain functionals, reliably interpreted as cell average diffusion coefficients. Finite-difference calculations with obtained in such a way full sets of cell characteristics can be then compared with benchmark MonteCarlo solution, providing reliable knowlege on quality of operative code under consideration. As an example of such study 1-, 2-, 4-, 8- and 12-group finite difference calculations of VVER model fuel assembly are compared with MonteCarlo solution. The quality of fictitious border cell transformation to the regular hexagon form was as well examined.

EXPERIENCE WITH VVER-440 FUEL OF THE 2-ND GENERATION WITH GD (AVERAGE ENRICHMENT 4.87%) AT SLOVAK NPP SITES
Sedláček M.1), Chrapčiak V. 1), Šimko J. 2), Greţďo O. 3)
1) VUJE, Inc., Slovakia
2) SE, Inc. – Enel, Mochovce NPP, Slovakia 3) SE, Inc. – Enel, Bohunice NPP, Slovakia
ABSTRACT
A new type of the 2-nd generation Gd fuel with average enrichment of 4.87 % has been used at Mochovce Units 1&2 since 2011 and at Bohunice Units 3&4 since 2012. Theoretical neutron-physical characteristics of reactor core are calculated by VUJE Inc. using the HELIOS and BIPR-7 programme complex. In the paper there are presented main features of the fuel and experience gained during the last 10 fuel cycles. Next, there is a comparison of theoretical and experimental neutron-physical characteristics and a discussion of the results. There are also presented some selected operational parameters such as relative power distributions (kq, kr, k0) measured by in-core monitoring systems. In the conclusion there is summarized experience with the 2-nd generation Gd fuel usage.

CALCULATIONS OF 3D FULL-SCALE FA MODELS WITH FEEDBACKS AND BURNUP USING MCU AND BIPR-7A CODES
Aleshin S.S., Bikeev A.S., Kalugin M.A., Kosourov E.K., Pavlovichev A.M., Sukhino- Khomenko E.A., Shcherenko Anna I., Shkarovskiy D.A
NRC “Kurchatov Institute”, Moscow, Russia
ABSTRACT
3D infinite lattices of full-scale VVER-1000 TVS-2M FAs with uranium fuel 4,4 % enrichment and uranium-erbium fuel 4,4 % enrichment and Er2O3 1 % wt are considered. FA burnup of FA working with the constant power level is modeled using MCU and BIPR-7A codes.
For different time moments effective multiplication factor and power distribution is obtained. Cases with and without power. with and without boron in water, with different control rods positions are considered. Results, obtained using BIPR-7A, are compared with MCU code results. Analysis of differences is conducted.

3D CALCULATIONS (WITHOUT FEEDBACK) OF EQUILIBRIUM VVER- 1000 FUEL CYCLE WITH THE REALISTIC BURNUP DISTRIBUTION IN CORE USING MCU AND BIPR-7A CODES
Bolshagin S.N., Kosourov E.K., Pavlovichev A.M., Prjanichnikov A.V., Shcherenko Anastasia I.
NRC “Kurchatov Institute”, Moscow, Russia
ABSTRACT
The emergency protection system efficiency and the volume reactor core power distribution for an equilibrium fuel cycle of VVER-1000 were studied. Calculations were performed with use of MCU and BIPR-7A program codes for 3D full-scale model of VVER- 1000 reactor without feedbacks with fuel burnup distributed in the reactor volume. Conclusions about deviations of the results obtained with use of the MCU program code from the results obtained by the BIPR-7A program code were drawn.

OUTCOMES OF CALCULATING CRITICAL ASSEMBLIES OF RIG ZR-6 USING SPECTRAL CODE TVS-M
Tsvetkov V.M.
NRC “Kurchatov Institute”, Moscow, Russia
ABSTRACT
The paper presents the results of ZR-6 critical assemblies calculations obtained with the spectral program TVS-M. The assemblies chosen were: K61, K91, K271, K331 and Xn, imitating the structure of a FA. Fuel assemblies contain water cavity, absorber rods of different composition including gadolinium. Also the fuel assemblies of cylindrical shape and some with the placement of several steel rods in reflector have been cosidered. Fuel assemblies have fuel rods of 9,1 mm in diameter and three different enrichments of 1,6%, 3,6%, 4,4% by 235U with lattice pitch 12,7 mm. The boron acid was used a set of experiments for achieving criticallity.
The results of effective multiplication factor calculations and the comparison of measured and computed activation of fuel rods are presented.

ANALYSIS OF APPLICATION OF VARIOUS VERSIONS OF FUEL ROD AXIAL PROFILING IN VVER-1000 USING THE SOFTWARE SAPFIR_95&RC
Zuev А.А., Ustinov А.N., Fateev М.V., Kurakin K.Yu. OKB “GIDROPRESS”
ABSTRACT
Now transition to one-and-a-half-year fuel cycle with power uprating up to 104 %Nnom comes to the end at Balakovo, Rostov and Kalinin NPP. The use of increased amount of fuel in the core and its enrichment means by the extended fuel cycles implemented at these Units, power uprating, this is provided due to application of FA with increased fuel stack: TVS-2М and TVSA-PLUS.
Experience in design and operation of such extended fuel cycles showed that the value of maximum linear power of fuel rods which is implemented in the upper part of the core comes near to the limiting design value and can reach it closer to the end of cycle. The revealed effect is of nonmonotonic power of FA during cycle because of non-uniform burnup of fuel in the core made up from FAs with a large (to 24-27 pcs) and, at the same time, a little (6 pcs) number of Gd-duel rods and becomes apparent, especially, in the cores, which comprise FA with extended fuel part.
Similar situation limits and complicates operation of Units and can be in some cases lead to power decrease. To avoid similar situations the specialists of RRC “Kurchatov Institute” proposed a method for decrease in linear heat rates of fuel rods by means of application of FA in axially-profiled Gd-fuel rods due to removal of gadolinium oxide from the end face parts of Gd-fuel rods or reduction of its content, this make possible, by the preliminary assessments, to decrease the maximum linear power of fuel rods within the range with minimum margin to the limiting design dependence and, thus, to increase this margin up to limiting design restrictions.
The paper represents detailed analysis of application of axial profiling of Gd-fuel rods by means of an example of equilibrium fuel loading of 18-month fuel cycle at Rostov NPP, Unit 2, by the results of the core neutronics calculations performed using the certified SP SAPFIR_95&RC_VVER. The activities aimed at optimization of Gd-fuel rod section height for profiling have been performed.
Separately, the computer code ReMi (ReadMicro) intended for processing of binary files (calculation results) of the RC code has been upgraded within the scope of the activity, this made possible to perform automatic processing of data on power distribution and burnup in the reactor core and substantially to accelerate the process of obtaining the final result

USAGE OF BURNT FUEL ISOTOPIC COMPOSITIONS, OBTAINED BY ENGINEERING CODES, IN MONTE-CARLO CODE CALCULATIONS
Aleshin S.S., Gorodkov S.S., Shcherenko A.I. NRC “Kurchatov institute”, Moscow, Russia
ABSTRACT
A burn-up calculation of large systems by Monte-Carlo code (MCU) is complex process and requires large computational costs. This fact makes Monte-Carlo codes usage complicated for project and operating calculations.
Previously prepared isotopic compositions are proposed to use for the Monte-Carlo code calculations of different system states with burnt fuel. Isotopic compositions are proposed to calculate by an approximation method. The approximation method is based on usage of a spectral functionality and reference isotopic compositions, that are calculated by engineering codes (TVS-M, PERMAK-A).
The multiplication factors and power distributions of FA are calculated in this work by the Monte-Carlo code MCU using earlier prepared isotopic compositions. The results of MCU calculations were compared with the results that were obtained by TVS-M code.

AER WORKING GROUP C ACTIVITIES IN 2014
Nemes
NPP Paks, Hungary
ABSTRACT

VVER OPERATION CONTROL MEASUREMENTS
Kalinushkin A.E., Milto N.V., Semchenkov Yu.M. NRC «Kurchatov Insitute», Moscow, Russia
ABSTRACT
In-core instrumentation system (ICIS) is primary means for control of VVER parameters related to safe operation. ICIS – the important part of the technical justification for safety and one of the elements of defence-in-depth principle realization. VVER development poses problems of ICIS improvement towards increased functionality, improvement of accuracy, reliability and performance. The modern ICIS is currently designed and implementing on a high capacity VVER reactors. The report presents the experience of the implementation and ongoing research for further improvement of ICIS for VVER reactors development.

THE SCORPIO-VVER CORE MONITORING AND SURVEILLANCE SYSTEM
FOR VVER-440 TYPE OF REACTORS
Jozef Molnár,
ÚJV Řeţ, a. s., Czech Republic
ABSTRACT
The SCORPIO-VVER core monitoring system has proved since the first installation at Dukovany NPP in 1999 to be a valuable tool for the reactor operators and reactor physicists. It is now installed on four units of Dukovany NPP (EDU, Czech Republic), on two units of Bohunice NPP (EBO, Slovak Republic) replacing the original Russian VK3 system and on the Full Scale Plant Training Simulator at the Centre for training and education of the reactor operators and reactor physicist in Trnava (SR). By both Czech and Slovak nuclear regulatory bodies the system was licensed as a Technical Specification Surveillance tool.
Since it’s first installation, the development of SCORPIO-VVER system continues along with the changes in VVER reactors operation. The system is being adapted according the utility needs and several notable improvements in physical modules of the system were introduced. The latest most significant changes were done in connection with implementation of a new digital I&C, loading of the optimized Gadolinium bearing Gd2M fuel assemblies, improvements in the area of core design (neutron physics, core thermal hydraulics and fuel thermal mechanics), adaptation of the system to up-rated unit conditions, in design and methodology of the limits and technical specifications checking (implementation of on-line SDM calculation) and improvements in the predictive part of the system (Strategy Generator). After the currently finished upgrades the SCORPIO-VVER monitoring system is still in focus of Central European NPPs with the roadmap of modification and implementation up to 2016. In October 2012 the Upgrade3 for the support of the new type of fuel with 4.87% enrichment of 235U loaded into the reactors at EBO Slovakia has been finished successfully. Thanks to the upgraded system all units could reach their 100% of nominal power.
During June 2013 the 6th Upgrade was started at EDU NPP in Czech Republic focused to the system enhancement and new function development up to 2016. The Dukovany’s Upgrade6 presents the most advanced upgrade in the system’s history. The SCORPIO-VVER system will be completely renewed, re-hosted to the new hardware, implemented the latest advanced calculation codes with enhanced accuracy and adapted to the local, national and international IEC standards and requirements too. The upgraded system will be put into the operation during the September 2014 on the 1st unit of EDU NPP.

EVALUATION OF EXPERIMENTAL UNCERTAINTY OF PHYSICAL MEASUREMENTS BASED ON START-UP DATA OF THE LATEST VVER-1000 PLANTS
Tsyganov S., Kravchenko A., Kraynov Yu., Nasedkin A., Chmykhun V. NRC “Kurchatov Institute”, Moscow, Russia
ABSTRACT
During 2000-th several new VVER-1000 units of different projects have been commissioned. At some of them, the similar initial fuel loadings were implemented; physical start up tests programme for measuring neutron-physics characteristics were also almost similar.
The paper presents the analysis of some result of neutron-physics characteristics were carried out during physics start up of VVER-1000 units that had similar initial fuel loading. The purposes of this analysis are an assessment of experimental uncertainties for chosen tests and refine, where it possible, the estimated experimental values of measured parameters.
The results of some tests that were carried out at units of Tianwan NPP (China), Bushehr NPP (Iran) and Kudankulam NPP (India) are under consideration. For each test it is discussed possible origins of experimental uncertainties and the confidence of made estimation.
The estimations presented in this paper may be useful for definition of criteria for the future physical test programme of new units. Refined experimental values of some parameters may be used for the purpose of code validation.

EXPERIMENTAL AND COMPUTATIONAL INVESTIGATIONS OF HEAT MASS TRANSFER INTENSIFIER GRIDS OF VARIOUS DESIGNS FOR EFFECTIVENESS
Kobzar L.L., Oleksyuk D.A., Semchenkov Y.M. NRC “Kurchatov Institute”, Moscow, Russia
ABSTRACT
The report describes the experimental work and calculations performed at the National Research Center “Kurchatov Institute” in recent years on the subject of heat mass transfer intensification.
Due to the increased power of operated blocks of NPP research of the intensification of heat-mass transfer held in various scientific and engineering organizations in Russia. In recent time many variations of intensifiers grids have offered. These grids have a positive influence, both the alignment parameters of the coolant for the assembly cross-section and to DNBR increase.
In NRC “Kurchatov Institute”, a large volume of experimental research on the determining the effectiveness different types of grids – intensifiers. Among the researching models was grid, which developed at NRC “Kurchatov Institute”. Results of experimental research can be used for verification of calculation models and codes.
The experimental data were obtained for the calculation of correlation coolant mixing and critical heat flux in rod bundles with intensifiers grids. Developed correlations have been introduced in subchannel code SC-INT. The comparison of the results of calculations by the code SC-INT with the experimental data is presented on the paper.

FINAL RESULTS OF THE PRIMARY SYSTEM HYDRAULIC CHARACTERISTICS AFTER MODIFICATION OF REACTOR COOLANT PUMPS` IMPELLER WHEELS AT BOHUNICE NPP EXECUTED BETWEEN 2012-2014
Martin Závodský, Jozef Hermanský Division for Preparation of NPP Operation
VUJE Inc., Okruţná 5, 918 64 Trnava, Slovakia Martin.Zavodsky@vuje.sk, Jozef.Hermansky@vuje.sk
ABSTRACT
The paper follows the similar papers presented last two years at AER Symposiums in Průhonice (Czech Republic) and Štrbské Pleso (Slovakia).
The project “Modification of reactor coolant pumps impeller wheels at EBO NPP” has been successfully finished in summer 2014 after the outage at Bohunice NPP Unit 3. Nowadays both Bohunice Units are operated without any difficulty with modified impeller wheels. All new impeller wheels, produced by CKBM in St. Petersburg, have the same diameter 595 mm. Measurement after final modification showed that the increase of coolant flow through the reactor at Unit 3 is circa 5,5 %. The same measurement executed last year showed that the increase of coolant flow at Unit 4 is circa 4,5 %. These values are even higher than expected values and values required by operator (Slovenské Elektrárne).
The paper summarizes the results of the primary system hydraulic characteristics measurement executed between 2012 – 2014 after partial modification (2 and 4 out of 6 impeller wheels) and after complete modification (6 out of 6 impeller wheels). Comparison of measured values with expected values is presented as well. The results of the thermal balances measurements are mentioned as well. The values of coolant flow through the reactor calculated by two different methodologies are in a good agreement. The results of the measurements demonstrated that the characteristics of reactor core cooling are improved.

NUMERICAL AND EXPERIMENTAL INVESTIGATION OF 3D COOLANT TEMPERATURE DISTRIBUTION IN THE HOT LEGS OF PRIMARY CIRCUIT OF REACTOR PLANT WITH WWER-1000
Saunin Yu., Dobrotvorski A., Semenikhin A., Ryasny S. JSC «ATOMTECHENERGO», Mytishchi, Russia
Kulish G., Abdullaev A.
Scientific-Technical Center “Nuclear fuel cycle”, Kharkov Physical-technical Institute, Kharkov, Ukraine
ABSTRACT
When implementing programs to increase capacity of the existing NPPs with WWER-1000 significantly increased demands on the accuracy of determining the weighted average reactor power depends strongly on the temperature stratification in the main circulation pipes, especially in the hot legs.
The paper presents the full-scale experimental data on the coolant temperature distribution in the hot legs obtained with help of regular I&C systems and first of all in-core monitoring system. These data was obtained at the pilot operation stage of unit-4 Kalinin NPP in a stationary condition at nominal power level. For the same condition the results of calculating simulation are presented under using the CFD code.
For the calculations of the temperature distribution in the main circulation pipes had used the computer code Solid Works Flow Simulation (COSMOSFloWorks). Simulation is carried out under the following conditions:

• calculation of the coolant velocity and temperature distribution, temperature condition of the reactor pressure vessel and its internals was performed by solving of the thermal conductivity dual problem;
• due to the small differences between loops coolant temperatures at the reactor inlet the stream swirling in the downcomer was not taken into account. Distribution of flow rates and coolant temperature at the core outlet was led to a 60° symmetry;
• in order to reduce the amount of calculations was simulated only “hot” part of primary
  circuit from the exit of coolant from the reactor core till turning pipe to steam generator.
  It was confirmed the presence of previously noted to all reactors common mechanism of formation stratified flows in the reactor vessel at the inlet of hot legs. According this mechanism the most heated jet of fluid and jet with an average temperature from the fuel assemblies at the central part of the core pass tube space up to perforation shell of the protective tube block without significant mixing and localized in the upper parts of the outlet nozzles. The coldest jet fuel assemblies from a number of peripheral and cold jets of leakage through the enclosure and annular gap localized at the bottom of outlet nozzles almost without mixing. Noted that the nature of localization is different in different loops with even and odd numbers.
  A satisfactory convergence of the experimental and calculated data was obtained. Noted some underestimation of the using the CFD code calculation of influence of perforation asymmetry in the protective tube block, and also noted the possible influence on the calculation results of exceptions of swirling in the downcomer of the reactor, to which impel the test results of the temperature field testing at the core inlet.
  Based on a comparative analysis of experimental and calculated data suggests possible solutions and the amount of further research.

THE PRACTICE OF CONSIDERING REACTIVITY SPATIAL EFFECTS WHEN MEASURING EMERGRNCY PROTECTION EFFICIENCY IN VVER REACTORS
Afanasiev D.V., Pinegin A.A. NRC “Kurchatov Institute”, Moscow, Russia Orlov V.I., Tolmasova E.V. VNIIAES, Moscow, Russia
ABSTRACT
The paper discusses the problem related to comparison of experimental and calculated values of emergency protection efficiency for VVER reactors. According to Regulations requirements measurements of emergency protection efficiency without of the one most effective control rod are carried out at the start of a fuel cycle. Search values of reactivity are turned out from indications of regular reactivity meters or using of personal computers by processing of currents of the ionization chambers (IC), located in biological shield channels, and further are compared to calculation results of the emergency protection efficiency made under the BIPR-7A program.
Drop of emergency protection into the reactor core is accompanied by sharp reduction of a neutron flux and by reduction of appearance speed of new generations of delayed neutrons emitters. As a result a new asymptotic (asymptotical) distribution of a neutron flux in the core isn’t established. In these conditions results of measurements depend on disposition IC and on local character of IC current formation. It leads to systematic methodical errors at determination of the emergency protection efficiency by reactivity meters indications, which are based on the turned solution of the dot kinetics equation (ORUK method). As a result we have considerable systematic differences between calculated and experimental values of reactivity because spatial effects are not taken into account.
For elimination of this systematic methodical error the modified ORUK (MOD ORUK) method was offered. This method, allowing account spatial reactivity effects, is implemented within the NOSTRA program. Under the offered method the value of emergency protection efficiency is accounted on the basis of measured IC currents by the ORUK method with use of special settlement amendments.
This paper presents results of processing by the MOD ORUK method of measurements of emergency protection efficiency for some of serial VVER-1000 reactors fuel charge and their comparison with settlement values. The analysis of the presented results shows that divergences between calculated and experimental values don’t exceed 15%.

PROCESSES OF ERROR DETECTION IN THE PATTERN OF CORE LOADING, COMMUTATION OF NITC AND MISALIGNMENT IN CONTROL ROD POSITION IN VVER-TOI REACTOR USING THE THEORY OF PATTERN RECOGNITION
Pinegin A.A., Ryzhov A.A.
NRC “Kurchatov Institute” Moscow, Russia
ABSTRACT
The presence of the extensive ICIS allows to detect violations in the reactor operation caused by improper loading of the core, misalignment in position of CR CPS working group, errors in commutation of KNI. Anomalies in the power distribution caused by the above reasons have some specific features. It allows to develop the problem-oriented procedure to identify these anomalies. These procedures can not be based on the recovered from ICIS data power distribution, because in considered cases procedures of power recovery are incorrect. As input information these procedures use indications of DCD.
The proposed algorithms are based on the estimation of linear functionals’ values using indications of DCD. Some parameters of the functionals (pattern of the functional) are determined by simulation of physical processes in the core. Another part of the functional parameters is determined by the methods of pattern recognition by minimizing losses due to errors in the recognition of the situation (Bayesian decision rule).
When determining the error in loading of the core Bayesian decision rules are calculated for each cell of the core within the measurement uncertainties, possible technological and computational errors.
When identifying a position of CR CPS as feature spaces it consider asymmetry in the power distribution in separate high-altitude layers of FA groups containing KNI and located near the considered CR CPS.
The report includes examples of numeric procedures that developed for detecting errors in loading the core and identifying positions of CR CPS.

AN ANALYSIS OF REACTIVITY PREDICTION DURING THE REACTOR START-UP PROCESS
Josef Bajgl
CEZ Inc., Dukovany NPP
Václav Krýsl, Jiří Švarný ŠKODA JS, a.s.
Czech Republic
ABSTRACT
The different VVER-440 core fuel loadings subcriticality is performed during the start- up process by boron dilution or control assembly withdrawn by macrocode MOBY-DICK calculations. The dynamic reactivity and quasicritical reactivity are compared and sensitivity of reactivity prediction at the low boundary of start–up interval has been provided on the bases of different modelling of ionization chamber (IC) response calculation. Special attention is paid to the impact of power distribution and spontaneous fission distribution form factor on IC response correction during control assembly movement. Precision and robustness of different corrections of IC signal processing in real core start-up processed was validated.

COMPARISON OF THE ACCURACY OF DIFFERENT METHODS OF MEASURING THE EFFECTIVENESS OF EMERGENCY PROTECTION SYSTEM IN VVER-1000
Afanasiev D.V., Pinegin A.A.
National Research Centre “Kurchatov Institute” Moscow, Russia
ABSTRACT
Measuring the effectiveness of the emergency protection system (EPS) in a VVER-1000 is one of the difficult tasks of start-up experiments. Usually experiments are carried out by the emergency protection reset while holding one, the most effective control rods above the core. Such experiments have a number of disadvantages. They, in particular, have significant methodological error. Systematic error of such experiments can be eliminated only by special methods involving computational programs. Another disadvantage of these experiments is their low sensitivity to deviations in multiplying properties of the central part of the core from the nominal values. This fact decreases informativeness of these experiments.
There is a method of determining the effectiveness of the emergency protection system by measuring the effectiveness of individual groups of control rods. Systematic error of this method is significantly less, but it requires some additional calculations.
Both methods are accurate in the absence of measurement errors, calculation errors, constant uncertainty, etc. The accuracy of determining the effectiveness of EPS using these methods has been investigated in the statistical model of experiments, created on the basis of the program NOSTRA. Statistical model of experiments, created on the basis of the program NOSTRA, takes into account the uncertainty of both the properties of the core, and experimental error. The report presents a comparative analysis of the accuracy of determining the EPS by both methods and informativeness value of the results.

AER WORKING GROUP D ON VVER SAFETY ANALYSIS – REPORT OF THE 2014 MEETING
S. Kliem
Helmholtz-Zentrum Dresden-Rossendorf Institute of Resource Ecology
P.O.B. 51 01 19, D-01314 Dresden, Germany S.Kliem@hzdr.de
ABSTRACT
The AER Working Group D on VVER reactor safety analysis held its 23th meeting in Garching, Germany, during the period 12-13 May, 2014. The meeting was hosted by the GRS Garching and was held in conjunction with the eighth workshop on the OECD Benchmark for Uncertainty Analysis in Best-Estimate Modelling (UAM) for Design, Operation and Safety Analysis of LWRs. Altogether 21 participants attended the meeting of the working group D, 19 from AER member organizations and 2 guests from non-member organization. The co- ordinator of the working group, Mr. S. Kliem, served as chairman of the meeting.
The meeting started with a general information exchange about the recent activities in the participating organizations.
The given presentations and the discussions can be attributed to the following topics:

• Safety analyses methods and results
• Code development and benchmarking including the calculation of the
  OECD/NEA Benchmark for the Kalinin-3 VVER-1000 NPP and 7th AER Dynamic Benchmark
• Future activities
  A list of the participants and a list of the handouts distributed at the meeting are attached
  to the report. The corresponding PDF-files of the handouts can be obtained from the chairman.

EXPERIENCES WITH DYN3D FOR SAFETY ANALYSES OF VVER440 REACTORS
Ctibor Strmenský
VUJE, Inc. Trnava, Slovakia ctibor.strmensky@vuje.sk
ABSTRACT
The history of using the dynamic code DYN3D in VUJE started 1994 and continued uptonow. The first version DYN3D code was version M2/M3 (up to 1999), next version H1.1 was used from 2001 to 2006 and present version 3.2 is used from 2006 upto now. DYN3Disdynamic code for pressurized reactors (PWR, VVER440/1000). DYN3D code is used for startup calculations and safety analysis.
In 2002 – 2004, we participated in international project VALCO, that was concerning to validation of dynamic and thermohydraulic codes and during 2010 – 2011, we participated in international project LWR-DEPUTY, that was concerning to development of the new combine-transmutation VVER-440 assembly based on the Mo matrix.
Preparation of input deck data is divided into several steps. At the start is necessary processing burnup data (BIPR7) with library of fuel data (HELIOS) into several blocks (cross-section data, feedback coefficients data, poison data (Xe, I, Sm, Pm), prompt and thermal neutron data, data of delay neutrons, neutron velocity data). All parameters are prepared for each fuel assembly. During finalisation process, all input data are formated into defined format. Several data areprepared in independent files (poison data, geometric and physics data for 59 assemblies, thermohydraulic data for whole core, hot channel data). One of key parameter is geometric data of the core. In past, we used 10 axial layers (2001 – 2005), where we obtained 3490 nodes (10*349). Next we extended to 20 axial layers model (2005 – 2010), where we obtained 6980 nodes (20*349). The present model contains 42 axial layers, that corresponds to 14658 nodes (42*349).
Second part of preprocessing is adjustment process, during input deck data are corrected by correction parameters (boron acid concetration, absorber data, albedo data and etc.).
The postprocessing part contains programs concerning to visualization data for 2D- graphs, 2D-cartograms or 3D-scene (stationary and transient).
Overview of using of DYN3D code in our institute is presented on 8 main interactive slides and 15 auxiliary interactive slides with figures and videos, which demonstrate preprocessing input deck data and postprocessing results.

COMPARISON OF THE RESULTS OF THE 7th DYNAMIC AER BENCHMARK – VVER-440 PRESSURE VESSEL COOLANT MIXING BY RE-CONNECTION OF AN ISOLATED LOOP
Kotsarev A.V., Lizorkin M.P.
NRC «Kurchatov Institute», Moscow, Russia
ABSTRACT
The 7th dynamic benchmark is a continuation of the efforts to validate systematically codes for the estimation of the transient behavior of VVER type nuclear power plants. This benchmark is a continuation of the work in the 6th dynamic benchmark. It is proposed to be simulated the transient – re-connection of an isolated circulating loop with low temperature in a VVER-440 plant. It is supposed to expand the benchmark to other cases when a different number of loops are in operation leading to different symmetric and asymmetric core boundary conditions. The purposes of the proposed benchmark are:

• 1) Best-estimate simulations of an transient with a coolant flow mixing in the Reactor Pressure Vessel (RPV) of VVER-440 plant by re-connection of one coolant loop to the several ones on operation,
• 2) Performing of code-to-code comparisons.
  Solutions were received from National Research Centre “Kurchatov Institute” (ATHLET/BIPR-VVER), UJV Rez (RELAP-3D) and HZDR (DYN3D/ATHLET). This paper gives an overview on the behaviour of the main thermohydraulic and neutron kinetic parameters in the provided solutions.

ANALYSIS OF CORE BEHAVIOR UNDER SEVERE ACCIDENTS USING SOCRAT/B1 CODE
Lityshev A., Pantyushin S., Aulova O., Gasparov D., Bukin N., Bykov M.
OKB “GIDROPRESS”, Podolsk, Russia pantyushin@grpress.podolsk.ru, tel:+7-4967-540479
ABSTRACT
Wide experience has been gained at OKB “GIDROPRESS” for twenty years in analyzing BDBA including severe accidents at VVER RP using the severe accident computer codes. Some foreign codes: MARCH, MELSIM (based on MELCOR 1.8.4), RELAP/SCDAPSIM – are available in the Computer Code Library at the enterprise. The Russian codes BISTRO and SOCRAT/В1 are also included into the library. Now the above-mentioned codes are a tool for investigations and design work aimed at justification of VVER RP safety. If in the nineties the majority of the work was performed using foreign computer codes, then for the last ten years a priority in execution of the work belonged to the Russian computer codes.
Analysis of in-core stage of accident with evaluation of occurrence of typical events, behavior of basic RP parameters, release of coolant mass and energy and core materials outside the reactor vessel (after damage of reactor vessel) is of the greatest interest for OKB “GIDROPRESS” as for the Chief Design organization of RP. Now the computer code SOCRAT/B1 certified in Rostekhnadzor and put into commercial operation at OKB “GIDROPRESS” is used to execute the computed simulation of BDBA with severe damage of the core.
Computer analysis of a wide spectrum of accidents is performed to justify safe operation of Units of VVER-1000 reactor plants (V-320). In connection with the accident at “Fukushima” Daiichi NPP a special attention is devoted to safety assessment of NPP with VVER-1000 RP under beyond design basis accidents including severe accidents.
The paper represents the results of calculation of core behavior under severe accidents using the computer code SOCRAT/В1.

NEUTRON-KINETIC AND THERMO-HYDRAULIC UNCERTAINTIES IN THE STUDY OF KALININ-3 BENCHMARK
Pasichnyk I1., Zwermann W1., Velkov K1., Nikonov S2.
1Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH, Garching, Germany {Ihor.Pasichnyk, Winfried.Zwermann, Kiril.Velkov}@grs.de
2VNIIAES, Moscow, Russia
niks.ki@mail.ru
ABSTRACT
The effects of nuclear data covariance on important reactor parameter are investigated. The analyses are performed on the base of the OECD/NEA coolant transient Benchmark (K-3) on measured data at Kalinin-3 Nuclear Power Plant (NPP).
For this purpose the GRS uncertainty and sensitivity software package XSUSA is applied to propagate uncertainties in nuclear data libraries to the full core coupled transient calculations. Moreover based on the previous thermo-hydraulic studies a set of most important thermo-hydraulic parameters is chosen and added to the uncertain input vector. A statistically representative set of coupled ATHLET-PARCS code calculations is analyzed and both integral and local output quantities are compared with the benchmark results of different participants as well as the measurements available in the benchmark. The work is a step forward in establishing a “best-estimate calculations in combination with performing uncertainty analysis” methodology for coupled full core calculations.

KALININ-3 BENCHMARK CALCULATION WITH COUPLED CODE ATHLET-PARCS
Pasichnyk I1., Nikonov S2., Velkov K1.
1Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH, Garching, Germany {Ihor.Pasichnyk, Kiril.Velkov}@grs.de
2VNIIAES, Moscow, Russia
niks.ki@mail.ru
ABSTRACT
The paper describes and discusses the new developed coupled neutronic/thermal- hydraulic code system based on the version of ATHLET 3.0 thermo-hydraulic code and the three-dimensional multi-group PARCS nodal kinetic code. The coupled code system is applied to analyze an asymmetric transient defined in the OECD/NEA coolant transient Benchmark (K-3) based on measured data at Kalinin-3 Nuclear Power Plant (NPP). The results of the calculation are compared with the set of detailed neutron flux and thermal- hydraulic measurements as well as with other benchmark participants’ calculations.
The main goal of the paper is the generation of a reference solution for further application in the uncertainty and sensitivity studies planned to be performed within the frame of the OECD/NEA UAM Benchmark.

COMPREHENSIVE ANALYSIS OF THE OECD/NEA KALININ-3 BENCHMARK (PHASE 1-3) BY USING AHLET AND ATHLET/KIKO3D COUPLED CODES
György Hegyi, András Keresztúri, István Trosztel, Zsolt Elter
Centre for Energy Research Hungarian Academy of Sciences, Budapest, Hungary
gyorgy.HEGYI@energia.mta.hu
ABSTRACT
The reliable prediction of the global and local reactor parameters is important issue to achieve a high validation stage of the coupled neutron-physics/thermal-hydraulics system codes which nowadays are considered to be the state of art by performing of accident analysis. It is similarly important to be able giving an answer and assurance of the following questions:

• how good and correct are the applied models in the simulation tools,
• the accuracy of the numerical methods incorporated in the codes,
• the completeness of the methodology to describe the multi-physical processes,
• the correctness of the required input data to model the processes.
  The paper gives a short summary of the Nuclear Energy Agency (NEA) of the
  Organization for Economic Cooperation and Development (OECD) VVER-1000 (Kalinin-3) ‘Switching-off of one Main Circulation Pump (MCP)’ transient benchmark for evaluating coupled system codes. The large number of available real plant experimental data made these benchmark problems very valuable.
  The experiment is very well documented, its uncertainties are known and the measurements were performed with a quite high frequency.
  In the first exercise instead of the complicate reactor-physical model the ATHLET 2.1 code with its point kinetics method (its parameters given in the benchmark specification) was used. Then in the second exercise, the core and the vessel are modelled. The inlet and outlet core transient boundary conditions and appropriate parameterized cross section data are provided by the benchmark team. In the third phase the full reactor is modelled.
  In the paper, our final solution of the above detailed benchmark problem is presented and the results are compared to documented experimental data.

SOLUTION OF THE OECD KALININ-3 BENCHMARK – EXERCISE 2 AND 3
Jan Hádek
ÚJV Rez, a. s., Hlavní 130, 250 68 Husinec-Rez, Czech Republic
jan.hadek@ujv.cz
ABSTRACT
The paper describes the results of the OECD benchmark problem that is connected with investigation of transient behavior in Kalinin-3 nuclear power plant (VVER-1000). The benchmark was defined by colleagues from VNIIAES (Russian Federation), RRC “Kurchatov Institute” (Russian Federation), GRS mbH (Germany) and Penn State University (USA). The benchmark simulation problem is the reactor coolant pump No. 1 switching off at the nominal power when the other three reactor coolant pumps are in operation during the experiment. Overall, the benchmark consists of four exercises. Exercise 1 describes the plant simulation with point kinetics model. Exercise 2 is devoted to the benchmark solution with using of stand-alone 3D reactor dynamic codes. The time-dependent thermal-hydraulic boundary conditions at the inlet of the reactor core are given in this task. Exercise 3 describes the solution of the problem by the coupled 3D reactor dynamic codes and the plant system codes. Exercise 4 is devoted to the uncertainty analysis. The paper is focused on solving of Exercises 2 and 3. Two different macroscopic cross section libraries are used. It is the library prepared in cooperation Penn State – Risk Engineering Ltd. (Bulgaria) and then the library from RRC “Kurchatov Institute”. The solution of Exercise 2 and 3 was performed by the code DYN3D and the coupled codes DYN3D/ATHLET. The calculation results obtained by both cross section libraries are compared with experimental values. The differences in the time behavior of compared parameters are discussed. In general, the results of benchmark calculations can serve as a reference when dealing with potential customers. They can also be used as supporting material during licensing of DYN3D and DYN3D/ATHLET codes in Czech Republic.

ANALYSIS OF FUEL THERMAL CONDITION BASED ON COUPLED NEUTRONIC AND THERMAL-PHYSICAL CALCULATIONS
Artemov V.G., Artemova L.M., Korotayev V.G., Miheyev P.A., Shemayev Yu.P. FSUE “Alexandrov NITI”, Sosnovy Bor, Russia
ABSTRACT
To assess conservatism in design calculations of fuel temperature condition for the VVER-type reactors, a numerical model has been developed and tested, which provides a subchannel calculation of fuel bundles using capabilities of the KORSAR thermal-hydraulic computer code and the SAPFIR_95&RC_VVER neutronic program package. The approach uses, first, results of power density subchannel simulation performed by the method of superposing the microscopic and macroscopic neutron flux distributions, tested on the VVER steady-state neutronic calculations during SAPFIR_95&RC_VVER verification, and, second, subchannel calculation of the thermal-hydraulic parameters of selected fuel bundles carried out with the use of the KORSAR code.
The following procedure is used for calculating accident scenarios in the proposed analysis:

• calculate the scenario in the complete-core approximation by using the KORSAR/GP code with 3D neutron kinetics; determine the hottest fuel bundles using the uncertainty and sensitivity analysis method; record the thermal-hydraulic and neutronic parameters of these fuel bundles as the scenario proceeds;
• reconstruct the subchannel power density distributions in the selected hottest fuel assemblies based on macroscopic neutron fluxes obtained from a dynamic calculation of the scenario and relative microscopic neutron flux distributions in fuel rods previously calculated for the selected fuel bundles;
• calculate the temperature condition of each individual rod using the KORSAR/GP subchannel model of a fuel bundle based on reconstructed subchannel power density distribution and thermal-hydraulic boundary conditions obtained from complete-core dynamic calculation of the scenario.
  Validity of the method for reconstructing power distribution in individual fuel rods is verified against results of test calculations performed by programs using the Monte Carlo method. The fuel bundle subchannel model is verified against results of experiments with electrically heated fuel bundles.
  The efficiency of the proposed approach is demonstrated by an example simulation of the VVER-1000 steam generator line break scenario

REACTOR DYNAMICS MODELS FOR SAFETY ANALYSIS IN AERB
Obaidurrahman K., Avinash J. Gaikwad
Nuclear Safety Analysis Division, Atomic Energy Regulatory Board (AERB), Anushaktinagar, Mumbai, India, 400094 (Corresponding author, e-mail: obaid@aerb.gov.in / obaid.iitb@gmail.com)
ABSTRACT
The nuclear power programme in India is being pursued with full regard for nuclear and radiation safety, which encompasses safety of plant personnel, public and the environment. The task of laying down necessary safety requirements and their enforcement are entrusted to the Atomic Energy Regulatory Board (AERB), which is the national authority to ensure that the use of ionizing radiation and nuclear energy in India does not cause undue risk to the health of workers and members of the public, and the environment. AERB is involved in safety review and licensing of variety of power reactors of different design and capacity. To provide technical support to its review process, AERB carries out independent safety analysis and research on selected important areas of nuclear and radiation safety. These safety studies help in taking important regulatory decisions during licensing process. Independent safety analysis requires a set of competent computational tools for which AERB has taken sincere initiatives to establish state-of-the-art best estimate analytical capabilities for analysis of core behavior, system thermal hydraulics, severe accidents and radiological impact assessment. Reactor dynamic is the integral part of any postulated initiating event (PIE) being analyzed. With induction of several types of reactors with diverse design in Indian nuclear map, it was imperative to develop detailed core modeling capability in AERB. Consistent with this objective a comprehensive 3D neutron kinetics code (TRIKIN) has been developed in AERB. TRIKIN model solves time dependent multigroup neutron diffusion to the highest level of approximation in full 3D core domain. This is achieved by flux factorization approach implemented by improved quasi static method. Finite difference method, assisted by a Chebychev’s acceleration scheme has been employed for spatial solution of neutron fluxes and Generalized Runge Kutta method has been used for time integration of amplitude function. Provision of triangular, hexagonal as well as square meshes in space solver renders high generality to the code, which allows its use for variety of reactors like PHWRs, VVERs, PWRs, BWRs and FBRs. For evaluation of core thermal state during evolving nuclear transient, a dedicated core thermal hydraulic model based on fuel pin simulation has been developed and internally coupled to the spatial kinetics model. Complete coupled TRIKIN model has been validated against a series of international benchmarks for VVER, PWRs and PHWRs. The present paper will describe salient features of AERB TRIKIN code and the glimpses of a few analytical studies carried out to support regulatory review process.

ANALYSIS OF HETEROGENOUS BORON DILUTION TRANSIENTS DURING OUTAGES WITH APROS 3D NODAL CORE MODEL
Kuopanportti, J.
Fortum Power and Heat Ltd, Espoo, Finland
jaakko.kuopanportti@fortum.com
ABSTRACT
A diluted water plug can form inside the primary coolant circuit if the coolant flow has stopped at least temporarily. The source of the clean water can be external or the fresh water can built up internally during boiling/condensing heat transfer mode, which can occur if the primary coolant inventory has decreased enough during an accident.
If the flow restarts in the stagnant primary loop, the diluted water plug can enter the reactor core. The flow can restart if a primary coolant pump is activated or if the natural circulation starts. In this paper, it is assumed that the natural circulation moves the plug into the core.
During outages after the fresh fuel has been loaded and the temperature of the coolant is low, the dilution potential is the highest because the critical boron concentration is at the maximum. The lid of the pressure vessel can be open or closed depending on the stage of the outage.
The idea of the analyses is to examine the behaviour of the core as clean or diluted water plugs of different sizes enter the core during outages. Therefore, the actual build-up of the diluted plugs was not analysed but the plugs were simply assumed to appear in the lower plenum or the downcomer in the beginning of the transient.
The analysis were performed with the Apros 3D nodal core model, which contains an own flow channel and 10 axial nodes for each fuel assembly. The wide-range cross section data was calculated with CASMO-4E.

ANALYSES OF BEYOND DESIGN BASIS ACCIDENT (BDBA) „HOMOGENEOUS” BORON DILUTION SCENARIOS
András Keresztúri, György Hegyi, Csaba Maráczy, István Trosztel
MTA EK, Centre for Energy Research, Hungarian Academy of Sciences Reactor Analysis Department
H-1525 Budapest 114, P.O. Box 49, Hungary andras.kereszturi@energia.mta.hu
ABSTRACT
Homogeneous boron dilution scenarios were analyzed by using the coupled KIKO3D- ATHLET code. The scenarios were named as homogeneous because of the very slow dilution caused by a rupture in a heat exchanger of the make up system. In case of the presented BDBA scenarios, it was supposed that after 30 minutes the operator doesn’t stop the dilution in spite of the prescriptions. Without the presented analyses, significant contribution of the „homogeneous” boron dilution to the CDF had to be assumed in the PSA analyses. The most relevant reactor states were certain shut down ones at low initial temperature and pressure. The analyses have shown that these events don’t lead to core degradation, although large amount of steam can be generated in the primary and secondary loops in absence of any reactivity protection. In spite of the very slow dilution, the time dependent boron concentration is becoming inhomogeneous. A reactivity initiated ATWS event occurs, where the reactor is becoming critical from time to time, demonstrating the strong self-regulating ability of the VVER-440 core also in case when the power rise is starting from low temperature and pressure. In spite of the dilution, the long term average of the power is stabilizing because of the adverse effects of the boron-free primary coolant discharge through the pressurizer relief valve and the decreasing moderator density.

APPLICATION OF STATISTICAL METHODS FOR SCRAM RELIABILITY ANALYSES
Kozlachkov A.N., Bykov M.A., Siryapin V.N., Shein V.P., Tribelev A.A. OKB “GIDROPRESS”, Podolsk, Russia
ABSTRACT
The problem of multiple control rods failures at scram working during the accident with a break of the main steam line was considered. Characteristic features of this accident are significant coolant temperature asymmetry in the core and danger of recriticality. Probability of scram functions non-fulfilment with a view to safety criterions was assessed.
Cases with different number of control rods failures were observed. Input data for multiple calculations was formed. Control rods failures were chosen at random.
Different statistical methods for post-processing were tested (by samples frequency, multidimentional linear regression, discriminant functional analysis). These methods can find out some laws.
The presented methodology allows to assess probability of failure in safety criterions for different number of control rods failures. It is a part of complete scram reliability analyses.

THE USE OF REALISTIC CODES
IN SOLVING THE PROBLEMS OF ANALYSIS OF REACTOR SAFETY
Semenovich O.V., Shaparau V.A., JIPNR‒Sosny of NASB, Belarus
ABSTRACT
System thermohydraulic codes ‒ software systems, designed for the simulation parameters of the coolant throughout the reactor facility. The coolant in pressurized water reactors during normal modes is a single phase liquid. In emergencies the coolant boils, turning into a vapor-liquid mixture. In the coolant implemented various modes of flow and heat transfer. Correctly describe such processes can only simulating vapor-liquid mixture in the approximation of a separate phase flow. System codes in which the coolant is described in the approximation of a separate phase flow have been called ‒ best estimate codes. Number of modern best estimate codes is small: about a dozen families. Here, the “family code” ‒ code and its modifications. The authors used a version of the software package ATHLET ‒ ATHLET/Mod2.2 (Cycle-A).
The authors thank the GRS (Germany) for licensed code ATHLET/Mod2.2(Cycle A) and for providing data for computational experiments.
The results of numerical simulation of the blowdown scenarios at the BC V-213 test facility, which are models of the hermetic compartments of NPP confinement (VVER- 440/213) with a full-scale fragment of bubble condenser, and all major equipment necessary for the test facility operation and testing, were presented in the report. It was shown how im- portant in simulation of the experimental facility to take into consideration the fragments of the “clipped” devices, which are located up to closed valves and, therefore, are parts (even a dead end) of circuit.

SELECTION OF FEEDBACK PARAMETER VALUES IN PREPARATION OF LOW-GROUP CONSTANT LIBRARIES FOR VARIOUS DYNAMIC CONDITIONS
Bryukhin V.V., Uvakin М.А., Kurakin K.Yu. OKB “GIDROPRESS”
ABSTRACT
The thermohydraulic computer code KORSAR/GP used for coupled neutronics and thermohydraulic calculations belongs to the best estimation code and is intended for simulation of normal operating conditions and accidents at reactor plants with VVER. The neutron kinetics module KARTA with the constant provision prepared by the software package SAPFIR_95&RC_VVER is used for calculation of core neutron kinetics of the VVER-type reactors in 3-D approximation in a set of KORSAR/GP.
Macroconstants of the reflector or core element depend on the values included into it, on thermal-physical parameters – feedback parameters (FBP). Dependence of the feedback parameter macroconstants in the module of spatial neutron kinetics is implemented based on multidimensional tables. Multidimensional table can be interpreted for each macroconstant as determination of its values in the parametric mesh knot points. Macroconstants for arbitrary values of feedback parameters are obtained by approximation of knot values of the macroconstants in the form of polynomials.
The present paper deals with the consideration of analysis of selection of quantity and values of the feedback parameter mesh knot points in preparation of low-group characteristic library. A method has been proposed for breakdown of the range of every possible values of feedback parameters into a sufficient quantity of intervals so that selection of any knot value of feedback parameter out of the range will not affect the results of neutronics characteristics when plotting the parametric mesh. Evaluation has been made for uncertainly of selection of the feedback parameter values on the results of safety parameters in simulation of various dynamic processes according to SP KORSAR/GP.

CFD SIMULATION OF THERMAL PROCESSES IN VVER REACTORS. COMPARISON WITH THE EXPERIMENTAL DATA
Bikezin S.K., Oleksyuk D.A.
NRC “Kurchatov Institute”, Moscow, Russia
ABSTRACT
This presentation describes the computational work performed by CFD-code STAR- CCM + for VVER and comparing the results with experimental data obtained from the National Research Center “Kurchatov Institute” in recent years on various topics.
Intensification of heat and mass transfer. In NRC “Kurchatov Institute”, a large volume of experimental studies on the effectiveness of intensifiers grids of different types. Among the studied lattice models was developed at NRC “Kurchatov Institute”. The experimental results obtained can be used for verifying CFD-code. Conversely, using the CFD-code the optimization of the geometry of the grid-intensifier to reduce the coefficient of hydraulic resistance.
The estimated impact of structural changes in the head on the coolant mixing in the area from out of the bundle of fuel elements to the location of the thermocouple. In order to improve the strength characteristics of fuel assembly (FA) VVER-440 third generation was upgraded design of the head. To unify the design of the head bore planned to be make for second generation FA. Since 2003 in the NRC “Kurchatov Institute” work was carried out to study the mixing of coolant in the head RK VVER-440. Were performed as computational studies on CFD-codes, subchannel code SC-1 and experimental studies on a full-scale mixing model of second generation FA. Experimental studies have allowed a verification of computational models and calculations based on the correction functions to get the thermocouple for a number of NPPs.

APPLICATION OF COMPUTATIONAL FLUID DYNAMICS IN NUCLEAR REACTOR SAFETY ANALYSIS
T. Höhne
Helmholtz-Zentrum Dresden – Rossendorf (HZDR) Bautzner Landstraße 400 | D-01328 Dresden Institute of Fluid Dynamics
Email: T.Hoehne@hzdr.de
ABSTRACT
The last decade has seen an increasing use of three-dimensional CFD codes to predict steady state and transient flows in nuclear reactors because a number of important phenomena such as pressurized thermal shocks, coolant mixing, and thermal striping cannot be predicted by traditional one-dimensional system codes with the required accuracy and spatial resolution. CFD codes contain models for simulating turbulence, heat transfer, multi-phase flows, and chemical reactions. Such models must be validated before they can be used with sufficient confidence in NRS applications. The necessary validation is performed by comparing model results against measured data. However, in order to obtain a reliable model assessment, CFD simulations for validation purposes must satisfy strict quality criteria given in the Best Practice Guidelines (BPG).
Numerical investigations on single phase coolant mixing in Pressurised Water Reactors (PWR) have been performed at the HZDR for almost a decade. The work is aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity.
On the other hand slug flow as a multiphase flow regime can occur in the cold legs of pressurized water reactors, for instance after a small break Loss of Coolant Accident (SB- LOCA). Slug flow is potentially hazardous to the structure of the system due to the strong oscillating pressure levels formed behind the liquid slugs. For the experimental investigation of horizontal two phase flows, different non pressurized channels and the TOPFLOW Hot Leg model in a pressure chamber was build and simulated with ANSYS CFX.
The behavior of insulation material released by a LOCA into the containment and the reactor core might compromise the long term emergency cooling systems. Subsequently, if the ECCS is operating in the sump recirculation mode, the debris suspended in the containment sump would begin to accumulate on the sump strainers. A small part could penetrate through the strainers and could be transported towards the reactor core. It was the aim of the numerical simulations to study where and how many mineral wool fibres are deposited at the upper spacer grid of a core.

IMPLEMENTATION OF CFD MODULE IN THE KORSAR THERMAL- HYDRAULIC SYSTEM CODE
Yudov Yu.V, Danilov I.G., and Chepilko S.S.
Alexandrov Research Institute of Technology (NITI), 188540 Sosnovy Bor, Leningrad region,Russia
Phone: (813-69) 22667, Fax: (813-69) 23672, E-mail: yudov@niti.ru
ABSTRACT
The Russian KORSAR/GP 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 1D two-fluid model for calculation of two-phase flows. Unsteady conservation equations are numerically solved using the semi-implicit scheme. The KORSAR/GP code includes a 3D reactor kinetics module. The code capability is intended for simulation of thermalhydraulic phenomena in circulation circuits of reactor systems and also for 3D modeling of coupled thermal-hydraulic (complete- core approximation) and neutronic processes in the reactor core.
A 3D CFD module is being developed by Alexandrov NITI for representing 3D effects in the downcomer and lower plenum during asymmetrical loop operation with the KORSAR/GP computer code. The CFD module uses Cartesian grid method with cut cell approach. Grids are generated using the adaptive refinement approach. The grid structure data is stored in Fully Threaded Tree (FTT) format. Small cut cells are merged with neighboring coarse cells. Convective terms are approximated using SDPUS-C1 high-resolution continuously differentiable scheme. Time integration of conservation equations is performed by the Kim-Choi second-order accurate implicit method. The Crank-Nicolson scheme is used together with linearization of non-linear terms. The Poisson equation for pressure at a new time level is solved by the multigrid method with Jacobi relaxation.
The paper presents a numerical algorithm for coupling 1D and 3D thermal-hydraulic modules in the KORSAR/GP code. To ensure mass and energy balances at the interface, convective fluxes for 1D module are calculated as a sum of fluxes at corresponding boundary edges of 3D module. Mass balance condition is used to couple pressures in the modules. A combined pressure field should be calculated at each time level. This is the most time- consuming operation. Cell sizes and, consequently, coefficients in discrete Poisson equations for pressure are different between 1D and 3D modeling domains by orders of magnitude causing slow convergence of iterative solutions. To increase the convergence rate, 1D module cells are included in the multigrid algorithm and considered as leaf cells on coarsest grid. Prolongation and restriction operators for these cells are trivial injection operators (values are merely copied as the level of the multigrid method changes). The relaxation procedure for 1D also uses simple Jacobi iterations.
The performance efficiency of the algorithm for coupling 1D and 3D modules was demonstrated by solving the benchmark problem of mixing cold and hot flows in a T- junction. The T-junction in that problem was modeled by CFD with prescribed inlet flow rates and fluid temperatures. The outlet part of the main channel downstream of the mixing region is modeled by the KORSAR/GP 1D module with given outlet pressures.

MODELING OF THERMOHYDRODYNAMIC PROCESSES IN FUEL ROD ASSEMBLIES OF WATER COOLED REACTORS
Semenovich O.V.
JIPNR‒Sosny of NAS of Belarus, Minsk, Belarus
ABSTRACT
The report discusses the results of computer simulation of thermohydrodynamic processes in the fuel rod assemblies of water cooled reactors.
The coolant in pressurized water reactors during normal modes is a single phase liquid. In emergencies the coolant boils, turning into a vapor-liquid mixture. In the coolant implemented various modes of flow and heat transfer. Correctly describe such processes can only simulating vapor-liquid mixture in the approximation of a separate phase flow.
To perform these investigations used software package that implements the three-fluid three-field subchannel mathematical model. Such a model is suitable for studying the processes of thermohydrodynamic in fuel assemblies in all possible modes of operation. Considered in detail the mathematical model and its discrete analogue. For the numerical implementation used the semi-implicit numerical scheme. All variables that are present in the source terms and the terms describing the interaction at the interfaces and surfaces of solid walls are treated in an implicit interpretation: estimated values inherent in the new (n+1) point in time. Also interpreted the phase velocities included in the convective terms of mass and energy, and pressure. Members of the convective flux of mass, momentum, energy, evaluated at the old (n) time step: explicitly. The “chess” grid is used for discretization in the spatial variable. A model of thermomechanical processes in rod fuel element that used in the subchannel code is considered.
The report analyzes the results of computer simulation

THERMO-HYDRAULIC EFFECTS OF VVER-1000 MIXING GRIDS. POSSIBILITY TO INCREASE THE ALLOWABLE ENERGY RELEASES
Shipov D.L., Samoilov O.B., Falkov A.A., Lukyanov V.E. JSC “Afrikantov OKBM”, Nizhny Novgorod, Russia
ABSTRACT
To increase the effectiveness of operating VVER-1000 and new VVER modules as applied to TVSA fuel assemblies, the plate-type mixing grid (MG) with mixing vanes without fuel element spacing function was developed.
Based on the experimental investigations, the geometry of plate-type MG was optimized from the viewpoint of hydraulic characteristics, mixing properties and turbulising influence on critical heat flux.
For the first time for the Russian fuel, the plate-type MGs were installed in 2008 at Kalinin NPP unit 1. Similar MGs in TVSA-T fuel assemblies were used as a part of the total core loads in 2010 and 2011 at Temelin NPP units 1 and 2.
The MG optimized as for mixing vane deflection angle, layout through the cross-section and layout in the upper spans between main spacer grids (SG) will be loaded in 2014 as a part of 48 TVSA-12PLUS fuel assemblies at Kalinin NPP unit 1.
Three plate-type MG in TVSA-12PLUS will ensure:
- small hydraulic loss (pressure loss coefficient 0.4);
- mixing coefficient increase by 5 times;
- increase of critical heat fluxes by 30-40% due to coolant flow turbulence.
The effectiveness of optimized plate-type MG is confirmed by the results of
experimental investigations of hydraulic characteristics at full-scale TVSA-12PLUS mockup with three MG; of mixing properties using large-scale 94-rod aerodynamic model and 19-rod electrically heated model with high power peaking; of MG turbulising properties influence on critical heat flux using 10 models with uniform and non-uniform radial and axial power distributions and with different MG and SG spans.
The thermo-hydraulic calculations for increased power VVER-1000 core and advanced with increased efficiency 3300 MW VVER core with TVSA-12PLUS with three MG and improved thermo-hydraulic characteristics were carried out.
The results show that TVSA-12PLUS with three MG implements safety criterion preventing DNB in case of increased fuel element power (F∆H=1.63) and maximum linear power density 448 W/cm and 380 W/cm at relative core height of 0.5 and 0.8 respectively.
Using of effective plate-type MG in TVSA-12PLUS together with DNB margin increase improves fuel element operation conditions owing to coolant heating levelling and mass steam quality decrease.

AER WORKING GROUP F ACTIVITIES IN 2014
Alekseev P.N., Dudnikov A.A.
NRC “Kurchatov Institute”, Kurchatov sq., 123182, Moscow, Russian Federation
Alekseev_PN@nrcki.ru ; Dudnikov_AA@nrcki.ru
ABSTRACT
Regular 15th session of the Working Group F «Nuclear Fuel Cycle Perspectives and Sustainability» was organized by National Research Centre “Kurchatov Institute” and held in Moscow, Russian Federation, July 14 – 17, 2014.
The themes of session were as follows:
1. Advancement of Nuclear Power structure for sustainable development 2. Future Nuclear Power resources supply for VVER
3. Advanced nuclear fuel cycles for VVER
4. REMIX-technology for VVER working in closed nuclear fuel cycle 5. Heavy nuclides using as burnable absorbers in VVER
6. Combined U-Pu-Th fuel cycle for VVER
7. Incineration of MA and transmutation of long-lived RW
8. Fuel cycle simulation
14 specialists from 4 countries have presented their reports:
1. Blandinsky V.Yu. (NRC KI, Russia) “Problems of Nuclear Power development as a
system”
2. Gurin A.V., Alekseev P.N. (NRC KI, Russia) “Research of the fuel component in
the cost of the electricity at the first stage of transition to the closed fuel cycle”
3. Bobrov E.A. et al. (NRC KI, Russia) “Perspective closed fuel cycles, based on
Regenerated Mixture Technology, combining use of thermal and fast reactors”
4. Brolly Á. et al. (Hungarian Academy of Sciences, Centre for Energy Research,
Hungary) “Development of the fuel cycle simulation code SITON”
5. Chibinyaev A.V. (NRC KI, Russia) “The main characteristics of the evolutionary
Super-VVER project (VVER-S) with spectrum shift regulation”
6. Chibinyaev A.V. (NRC KI, Russia) “The main characteristics of the innovative
Super-VVER project”
7. Mikisek M. (ÚJV Řeţ, a. s., Czech Republic) “Applicability of Small Reactors in
the Czech Republic”
8. Frybort J. (presented by Mikisek M.) (ÚJV Řeţ, a. s., Czech Republic) “Thorium
Fuel in VVER-1000 Reactor”
9. Szabó F. (MVM Paks II. Ltd., Hungary) “Refuelling cycle strategy comparison of
AES-2006”
10. Bolshakov V.V. et al. (NRC KI, Russia) “Research of the potential of thorium fuel
for vver-1000”
11. Shmelev A.H., Kulikov G.G. (NRNU Moscow Engineering Physics Institute,
Russia) “Improvement of LWR fuel cycle: Potential role of fusion hybrid with Th- blanket”
12. Dařílek P. (VUJE a.s. , Trnava, Slovakia) “ALLEGRO development status”
13. Tsibulsky S.V., Davidenko V.D. (NRC KI, Russia) “The effective way of burning power plutonium in reactors VVER”
14. Nevinitsa V.A., Dudnikov A.A. et al. (NRC KI, Russia) “Prospects of subcritical molten salt reactor for minor actinides incineration in closed fuel cycle”

PROSPECTS OF SUBCRITICAL MOLTEN SALT REACTOR FOR MINOR ACTINIDES INCINERATION IN CLOSED FUEL CYCLE
Alekseev P.N., Blandinsky V.Yu., Dudnikov A.A., Fomichenko P.A., Nevinitsa V.A., Frolov A.A., Lubina A.S., Sedov A.A., Subbotin A.S.
NRC “Kurchatov Institute”, 123182, Moscow, Kurchatov sq. 1, Russian Federation
ABSTRACT
The subcritical molten salt reactor (SMSR) with fuel composition of heavy nuclide fluorides in molten FLiNaK salt and with external neutron source, based on 1 GeV proton accelerator and molten salt cooled tungsten target is considered. SMSR is proposed for minor actinides (separated from spent fuel VVER-1000 light water reactor) incineration and for 233U conversion from 232Th. Paper presents the results of parametrical analysis of equilibrium nuclide composition of molten salt reactor with minor actinides feed in dependence of core dimensions, average neutron flux and external neutron source intensity. Reactor design is defined; requirements to external neutron source are posed; heavy nuclides equilibrium and fuel cycle main parameters are calculated.

ALLEGRO PROJECT
Branislav HATALA, Petr DARILEK,
VUJE, a.s., Okruzna 5, 918 64 Trnava, Slovak Republic;
Branislav.Hatala@vuje.sk
ABSTRACT
The goal of the ALLEGRO Project is to design, build and operate the first Gas cooled Fast Reactor (GFR) Demonstrator. The GFR belongs to the 4th generation of nuclear reactors. Prototype of the gas cooled reactor ALLEGRO is a subject of common interest of association of countries France, Slovakia, the Czech Republic and Hungary with the goal to deploy prototype of the reactor in Central Europe and present its functionality. The four Visegrad countries (V4) agreed on establishment of “V4G4 Centre of Excellence”, for performing common R&D in the area of Generation IV reactors. The agreement allows participants to join their resources for development of Generation IV reactors and thus contribute to meeting the goal of sustainability of nuclear energy while maintaining the highest safety standards.

INVENTORY AND DECAY HEAT CALCULATIONS FOR 4.0 % VVER-440 FUEL ASSEMBLY USING THE CODES SERPENT AND SNF
Tuukka Lahtinen
Fortum Power and Heat Ltd, Espoo, Finland
tuukka.lahtinen@fortum.com
ABSTRACT
As accurate as possible computational information on decay heat power of irradiated nuclear fuel is of essential importance in various applications at NPP. The approach of not only examining the decay heat power as an integral quantity but also modeling the composition of the fuel elaborately on a nuclide-specific level provides valuable information which is, in addition to pure decay heat analyses, also applicable in various other areas.
In this study, a single 4.0 % enriched VVER-440 fuel assembly is modeled in a 2D geometry. The assembly is irradiated up to 50 MWd/kgU burn-up using a simplified irradiation history by keeping the parameter values of the fuel and coolant properties constant. The study aims at comparison of the decay heat powers and nuclide inventories resulting from the simulations that are carried out using the Serpent and SNF codes.
In addition to comparison of results in the nominal conditions case, also variation cases are examined. In these cases, irradiation parameter values are deviated from the nominal ones and the impact on the time-dependent decay heat power on a nuclide-level is examined.
The current version of the SNF code is capable of modeling 357 radionuclides that are important decay heat contributors at least during some limited decay-time period. On the other hand, the nuclear data libraries applied in the Serpent Monte Carlo code, include neutron interaction data for over 400 and radioactive decay data for over 3800 nuclides. The comparison of the results between the two codes provide data which, among other things, can be used to verify that the full depletion calculation chain behind the utilization of the SNF code, is capable to model all the relevant decay heat nuclides.

COMPREHENSIVE RESEARCH OF THE PRACTICAL REALISATION OF THE THORIUM NUCLEAR FUEL CYCLE
Bolshakov V.V.
NRC “Kurchatov Institute”, Moscow, Russia
Bashkirtsev S.M., Morozov A.G. Lightbridge Corp. , Virginia, USA
ABSTRACT
A set of works is completed for the practical realization of the thorium nuclear fuel cycle. Two-zone Thorium FA was designed in form-factor of FA for VVER-1000 reactor. Design was optimized by a series of preliminary neutronics and thermo-hydraulic calculations. The peripheral zone of the FA is a blanket-zone. It contains cylindrical FE a little bit smaller in diameter than the traditional FE. Central seed-zone of the FA is formed by FE twisted along the longitudinal axis. These special FEs have triple-vane cross sections and are self-arranged in hexagonal lattice. Success in production of twisted FEs by the extrusion technique allows their using in FA design. Short samples of the FE both types of the Thorium FA underwent ampule tests at the research reactor. A series of tests were performed at the thermo-hydraulic test facility KS in the NRC “Kurchatov Institute”. Pressure drop and CHF investigations were completed for the bundles of the full scale FE simulator. Methods and techniques obtained from the set of works have additional advantages.

MODELS OF MULTI-CRITERIA ANALYSIS AND CONTROL OF PROPERTIES OF DISTRIBUTED OBJECT SYSTEM OF CLOSED NUCLEAR FUEL CYCLE
Ignatov A.A., Kormilitsin M.V., Subbotin S.A., Alekseev P.N., Bulanova T.M. NRC “Kurchatov Institute”, Moscow, Russia
ABSTRACT
Formed complex technical and economic models for quantitative description (VEA) of integrated indicators and local operating characteristics of a distributed system of production units (nodes) object lifecycle closed nuclear fuel cycle (CNFC). The proposed model takes into account all the major blocks of two conjugate process stages of nuclear reactors “thermal- fast” neutrons. For each stage of a typical nuclear power accounted for interacting structure of process nodes (units): 1) the production of fuels (fuel rods), 2) reactor zone, including how converters nuclear energy into electricity, and 3) regeneration of spent nuclear fuel materials, 4) treatment radioactive waste management, 5) decommissioning. relevant complicated subsystems represented closed multilevel (hierarchical) information-analytical model with multiple feedback loops that account (status characteristics) production units. Given components: capital investment costs, taxes, and maintenance costs, depreciation, discounting the fuel-costs-parties costs of energy production facilities. The proposed analytical system provides optimization design performance management subsystems, particularly in aspects of factors influence the market environment, limited financial resources, etc.
One of the primary applications performance management process is CNFC offers for substantiation of the design decisions in relation created where FSUE-«Zheleznogorsk» (Krasnoyarsk)- Experimental Demonstration Center (ODC), the main requirement to which is to develop the complex physical and chemical processes of regeneration of fissile material for spent nuclear fuel. In this fractionation processes radionuclides spent fuel must address the following list of applications:

• Allocation of fissile nuclides for use in nuclear power.;
• testing of new technologies and equipment, reducing processing costs;
• Issuance of initial data for design and construction. large-scale plant for the
  reprocessing of spent nuclear fuel.
  Another subtask report – justification expert methods practical justification (use) in the
  international markets (ICFS) fuel services production of nuclear energy (energy production, storage volumes and fuels) as analytical framework SWOT-analysis technology competitiveness CNFC as complex properties "opportunities-threats -prospects. Proposed analytical subsystems are also interesting as a tool of learning in relation to the characteristic properties of innovative scenarios of nuclear energy.

CORRELATION DEPENDENCIES FOR DETERMINATION OF BURNUP AND ACTINIDES CONTENT IN SPENT NUCLEAR FUEL
Zhemzhurov M.L., Serebriany G.S., Rudovich D.O. JIPNR– Sosny, of NAS of Belarus, Minsk, Belarus
ABSTRACT
Correlation dependencies have been obtained to determine the burn-out and content of actinides in spent nuclear fuel based on an analysis of published experimental data on the definition of burnout and isotopic composition of spent nuclear fuel rods for samples of PWR and VVER reactor types using statistical methods. Given dependencies of the mass concentration of actinides at the end of irradiation after checking for normality correspond to the initial experimental data. These relationships may be of interest to use for non-destructive methods for the purpose of express evaluation of spent nuclear fuel burnout at all stages of the nuclear cycle.

SOME HELIOS-2 CALCULATIONS IN ALLEGRO PROJECT
Radoslav ZAJAC, Petr DAŘÍLEK
VUJE, Inc., Okruzna 5, 918 64 Trnava, Slovakia; radoslav.zajac@vuje.sk,
petr.darilek@vuje.sk
ABSTRACT
Introduction of the article is focused on a short ALLEGRO core characterization and fuel type description. Calculations of ALLEGRO subassemblies by HELIOS 2.1 code is shown. Three fuel sub-assemblies were modelled in infinite lattice by HELIOS 2.1 – starting MOX assembly, ceramic UPuC sub-assembly and ceramic UPuC sub-assembly with thermal insulant. The super- cell model of MOX core with experimental ceramic sub-assemblies is listed at the end of this presentation and it consists of six MOX sub-assemblies and one ceramic UPuC sub-assembly situated in the centre. Multiplication factor for infinite lattice kinf was followed in all of calculated cases. All of analyses in the paper were inspired by an active participation at GoFastR project – component of FP7 EU.