Evaluation of Fuel Assembly Coolant Outlet Temperatures of Kalinin-3 Benchmark
25th Symposium of AER on VVER Reactor Physics and Reactor Safety (2015, Balatongyörök, Hungary)
Reactor physics experiments and code validation (benchmarks)
Abstract
EVALUATION OF FUEL ASSEMBLY COOLANT OUTLET TEMPERATURES OF KALININ-3 BENCHMARK Jan HdekJV Rez, a. s., Hlavn?? 130, 250 68 Husinec-Rez, Czech Republicjan.hadek@ujv.czABSTRACT The OECD Kalinin-3 benchmark problem is connected with investigation of transient behavior in Kalinin-3 nuclear power plant (VVER-1000). 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. Two different macroscopic cross section libraries are used. The paper is focused on evaluation of fuel assembly coolant outlet temperatures received during the solution of Exercises 2 and 3. 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 between calculated and measured fuel assembly coolant outlet temperatures are discussed. The results of similar comparison from Temeln NPP are briefly presented.1. INTRODUCTIONThe OECD benchmark Kalinin-3 (VVER-1000) benchmark was defined as a reactor coolant pump No. 1 switching off at the nominal power when the other three reactor coolant pumps are in operation during the experiment 1. The basic reactor core geometric and material parameters were given. The Penn State – Risk Engineering Ltd. (PSU-REL) macroscopic cross section library was distributed by benchmark organizers. Using of this library was highly recommended by benchmark team. As a second option, the macroscopic cross section library created at the Kurchatov Institute (KI) was used. The benchmark participants have to use also their own plant input data decks developed according to the needs of their own VVER-1000 NPP system codes. ??JV Rez, a. s. has been participated in solving of Exercises 1 to 3 of this benchmark. Some results of Kalinin-3 benchmark were presented at the 24th Symposium of AER in Sochi 2. In this paper, the main attention is focused on the comparison of the calculated and experimental fuel assemblies coolant outlet temperatures. It may also be understood as a contribution to the validation of three-dimensional reactor dynamic code DYN3D and its coupling with thermal-hydraulic system code ATHLET.EXERCISE 2Exercise 2 is focused on the evaluation of the three-dimensional reactor dynamic core model response 1. The time-dependent boundary condition at the inlet and outlet of reactor core were given by benchmark organizers. They were prepared on the basis of coupled BIPR/ATHLET code calculation. Including of Exercise 2 to the analysis of fuel assemblies coolant outlet temperature has its own importance. Using uniform time-dependent boundary conditions, we can compare Exercise 2 results with the results of the next Exercise 3 and, if necessary, we can fine tune our Exercise 3 input deck.2.1 Computer code usedThe reactor dynamic code DYN3D is used as a tool for Exercise 2 calculation. DYN3D is a 3-D core model for dynamic and depletion calculations in LWR cores with hexagonal or quadratic fuel assembly geometry 3, 4 developed by Helmholtz Zentrum Dresden ? Rossendorf (HZDR). The neutron-kinetic model is based on the solution of the 3-D two-group neutron diffusion equation by nodal expansion methods. Different methods are used for hexagonal and quadratic fuel assembly geometry. It is assumed that the macroscopic cross sections are spatially constant in a node being the part of a fuel assembly. In the case of hexagonal fuel assemblies, the diffusion equation in the node is transformed into a 2-D equation in the hexagonal plane and 1-D equation in the axial direction. The two equations are coupled by the transverse leakage terms, which are approximated by polynomials up to the second order. Considering the 2-D equation in the hexagonal-plane, the side averaged values (HEXNEM1) or the side averaged and the corner point values (HEXNEM2) of flux and current are used for the approximate solution of the diffusion equation. In the case of Cartesian geometry, the 3-D diffusion equation of each node is transformed into 1-D equations in each direction x, y, z by transverse integrations. The equations are coupled by the transverse leakage term. In each energy group, the 1-D equations are solved with the help of flux expansions in polynomials up to second order and exponential functions being the solutions of the homogeneous equation. The fission source in the fast group and the scattering source in the thermal group as well as the leakage terms are approximated by the polynomials. In the steady state, the homogeneous eigenvalue problem or the heterogeneous problem with a given source is solved. Inner and outer iteration strategy is applied. The outer iteration (fission source iteration) is accelerated by Chebychev extrapolation. The steady-state iteration technique is applied for the calculation of the initial critical state, the depletion calculations, and the Xe and Sm dynamics. Concerning reactivity transients an implicit difference scheme with exponential transformation is used for the time integration over the neutronic time step. The exponents in each node are calculated from the previous time step during the iteration process. The precursor equations are analytically solved, assuming that the fission rate behaves exponentially over the time step. The heterogeneous equations obtained for each step are solved by inner and outer iteration techniques similar to the steady state.In addition, the thermal-hydraulic model of the reactor core and the fuel rod model are implemented in the FLOCAL module 4, which is a part of DYN3D. The reactor core is modeled by parallel coolant channels, which are coupled hydraulically by the condition of equal pressure drop over all core channels. Thermal-hydraulic boundary conditions for the core like coolant inlet temperature, pressure, and coolant mass flow rate or pressure drop must be given as input for DYN3D. FLOCAL comprises a one- or two-phase coolant flow model on the basis of four differential balance equations for mass, energy, and momentum of two-phase mixture and the mass balance for the vapor phase allowing the description of thermodynamic nonequilibrium between the phases, a heat transfer regime map from one-phase liquid up to postcritical heat transfer regimes an superheated steam. A fuel rod model for the calculation of fuel and cladding temperatures is implemented. A thermomechanical fuel rod model allows the estimation of relevant heat transfer behavior of the gas gap during transients and the determination of some parameters for fuel rod failure estimation.2.2 Reactor core modelThe reactor core model is fully described by the code DYN3D. It consists of 163 parallel coolant channels where each from them corresponds to one fuel assembly. The reactor core is surrounded by the band of artificial reflector assemblies. It means that it consists from 211 assemblies. The reactor core is divided into 12 axial layers (one layer for lower axial reflector, 10 layers for heated part of the core, one layer for upper axial reflector). The zero flux boundary conditions are applied for outer surface of axial reflector. Figure 1 shows the numbering of fuel assemblies in DYN3D core model in 60 degree rotational and full core symmetry and also the distribution of materials contained in PSU-REL cross section library. Locations of control rods (control clusters) in reactor core model are shown in Figure 2.Fig. 1 Numbering of fuel assemblies and their typesFig. 2 Numbering of fuel assemblies and location of control rods2.3 Cross section librariesTwo-group homogenized macroscopic cross sections and kinetic parameters for composition of each assembly were provided in nemtab format which was used also in previous OECD/NEA benchmark V1000-CT1 5. Two sets of tables were available. One for unrodded (nemtab) and other for rodded assembly nodal compositions (nemtabr). Two types of nemtabr cross section data (for dysprosium and boron absorbers) were given for fuel assemblies with inserted control clusters during the transient. Reflector compositions were defined for lower and upper reflector and for one radial reflector. The parameterization of macroscopic cross sections and kinetic parameters within the nemtab format (for fuel temperature, moderator density and moderator temperature) was done in three-dimensional space 1. The burn-up was taken into account in the composition of each axial node.Two types of cross-section libraries were provided. The first library was prepared on the basis of cooperation between Penn State and Risk Engineering Ltd. (PSU-REL X-S) 6. It contains a total of 973 materials in nemtab format. The PSU-REL X-S library takes into account the symmetry of materials arrangement in the reactor core model. The second library was prepared at the Kurchatov Institute (KI X-S) 7. It contains a total of 1761 materials and it is prepared in a different nemtab format. This library was prepared for the whole reactor core model.2.5 HFP initial steady-state conditionsThe HFP initial steady-state conditions are given in Table 1. Table 1 Exercise 2 ? HFP initial stead-state conditionsParameterValueInitial reactor core parametersReactor power, [MW]2965Cycle status, [EFPD]130.6Primary pressure, [MPa]15.52Reactor mass flow rate, [m3/h]88073Moderator inlet temperature [K], ( [oC])560.94 (287.79)Boron acid concentration [g H3BO3/kg H2O]3.60Control rod group 1-9 position [cm]369.25Control rod group 10 position [cm]309.252.6 Time dependent boundary conditionsThe time dependent boundary conditions were provided by benchmark organizers. They are shown in Figures 3 to 6. Fuel assembly inlet mass flow rates, fuel assembly inlet temperatures were obtained on the basis of calculations with coupled code BIPR/ATHLET 1. Core outlet pressure and control rod positions were taken from experimental data. Fig. 3 Fuel assembly inlet mass flow ratesFig. 4 Fuel assembly inlet temperaturesFig. 5 Core outlet pressureFig. 6 Control rod positions2.7 Calculated vs experimental resultsComparison of calculated and experimental nuclear powers can be seen in Figure 7. The calculated time-dependent nuclear powers are at a lower level in comparison with experiment. This is valid for both used cross-section libraries; simultaneously the KI X-S library gives a higher level of nuclear power at the end of transient. Figures 8 to 15 show the comparison of calculated as well as experimental fuel assembly coolant outlet temperatures. The PSU-REL X-S library gives higher fuel assembly coolant outlet temperatures at the beginning of transient. At the end of transient, we can observe the opposite phenomenon where fuel assembly coolant outlet temperatures received with KI X-S library are higher. It can be seen in Figures 12 to 15. However, we can say that the calculated temperatures obtained with using of both libraries are in reasonable agreement. Another observed phenomenon is that the calculation results are systematically overestimated in comparison with experiment. The maximum deviation is about 10 oC at the beginning of transient. The minimum deviation is approximately 2 oC when the control rods of group No. 9 and 10 are in lower positions. It is shown in Figures 12 and 14. Fig. 7 Nuclear powersFig. 8 Fuel assembly coolant outlet temperature PSU-REL X-SFig. 9 Fuel assembly coolant outlet temperature KI X-SFig. 10 Fuel assembly coolant outlet temperature KI X-S vs PSU-REL X-SFig. 11 Fuel assembly coolant outlet temperature KI X-S vs PSU-REL X-S vs experimentFig. 12 Fuel assembly coolant outlet temperature t = 0 sFig. 13 Fuel assembly coolant outlet temperature t = 45 sFig. 14 Fuel assembly coolant outlet temperature t = 71 sFig. 15 Fuel assembly coolant outlet temperature t = 300 sEXERCISE 3Exercise 3 is focused on the best-estimate coupled code plant transient modeling 1. Using of three-dimensional reactor dynamic code coupled with system plant thermal-hydraulic code enables fully simulate the system plant response.3.1 Computer code usedThe coupled code DYN3D/ATHLET is used as a tool for solution of Exercise 3. The thermal-hydraulic computer code ATHLET (Analysis of Thermal Hydraulics of LEaks and Transients) was developed by the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH for the analysis of anticipated and abnormal plant transients, small and intermediate leaks as well as large breaks in LWR 8. The main code features are advanced thermal hydraulic, modular code architecture, separation between physical models and numerical methods, preprocessing and postprocessing tools, and portability. ATHLET offers the possibility of choosing between models for the simulation of fluid dynamics. The first option is five-equation model with separate conservation equations for liquid and vapor mass and energy, and a mixture momentum equation, accounting for the thermal and mechanical nonequilibrium, and including a mixture level tracking capability. Further, a complete two-fluid model with separate conservation equations for liquid and vapor mass, energy, and momentum is available. Additional mass conservation equation can be included for the description of boron transport within a coolant system. The time integration of the thermal fluid dynamics is performed with a general solver for ordinary differential equations called FEBE (Forward-Euler, Backward-Euler).For the simulation of heat conduction, a one-dimensional (1-D) heat conductor module is available. This model can simulate the 1-D temperature profile and heat conduction in plates, hollow and full cylinders in radial direction. A heat transfer package that covers a wide range of single-phase-flow and two-phase-flow conditions is available inside the module. Correlations for critical heat flux and minimum film boiling temperature are included; evaporation and condensation directly at heating surface are calculated. A quench front model for bottom and top reflooding is also available.The code development is accompanied by a systematic and comprehensive validation program. A large number of integral experiments and separate effects tests have been calculated by GRS and by independent organizations. The range of applicability has been extended to the Russian reactor types VVER and RBMK in cooperation with foreign partner organizations. ATHLET is being applied by numerous institutions in Germany and abroad.The model of external coupling of the 3-D reactor dynamic code DYN3D with the system code ATHLET was used 9. The whole reactor core model is removed from the ATHLET plant model and the core is completely substituted by the DYN3D core model. The thermal hydraulics of the whole NPP is split into two parts: the first part describes the thermal-hydraulic of the reactor and the second solves the thermal-hydraulics of the rest of the NPP system. The interfaces between them are located at the bottom and the top of the core. The pressures, mass flow rates, enthalpies, and boron acid concentrations are transferred at the interfaces. The parameters exchange is performed by the GCSM (General Control Simulation Module) of system code ATHLET 8. The SATM (Self Adapting Turbulent Mixing) model for lower and upper plenum was implemented in the DYN3D code 10. 3.2 Reactor core model and nodalization schemesThe reactor core model is identical to the model used for Exercise 2. Locations of coolant channels in reactor core model and their connection with primary system coolant loops are shown in Figure 16. The standard ATHLET input deck for VVER-1000 was used. All four loops of reactor coolant system are fully modeled as well as the pressurizer system. The pressurizer surge line is connected to the hot leg of the fourth loop; the spray line is connected with the cold leg of the first loop. The reactor upper plenum was modeled on three levels with four parallel volumes with cross-connections. The space under the reactor pressure vessel head was created by one volume. The lower plenum was modeled as four parallel channels. The core bypass was simulated by the DYN3D code.The secondary circuit modeling corresponds to the primary circuit. There are fully modeled steam generator (SG) secondary side, main steam system (main steam lines (MSL), main steam header (MSH)), and feedwater (FW) system lines. The simplified scheme of coolant loop No. 1 nodalization is given in Figure 17.1762696843166Hot leg00Hot leg992198828218Hot leg00Hot leg7848602071847Lower plenum00Lower plenum7372351278255Upper plenum00Upper plenum329608392014Pressurizer00Pressurizer1068121659078Pressurizer surge line00Pressurizer surge line25280131087650MCP00MCP18564971595120Downcommer00Downcommer2391985130036Steam generator00Steam generator9399041745615Core bypass00Core bypass14255051565813Core 00Core 27203401854835Cold leg00Cold legFig. 16 Fuel assemblies numbering and location of coolant loopsFig. 17 Reactor vessel and primary circuit nodalizationTwo-group homogenized macroscopic cross-section libraries are identical to those that are used in Exercise 2 – i.e. PSU-REL X-S and KI X-S libraries.3.4 HFP initial steady-state conditionsThe HFP initial steady-state conditions are given in Table 1. 3.5 Time dependent boundary conditionsThe time dependent boundary conditions for control rod group No. 9 and No. 10 positions, main steam header pressure, feedwater mass flow rates and temperatures were taken from experimental data. They are shown in Figure 6 and in Figures 18 to 20. Fig. 18 Main steam header pressureFig. 19 Steam generator feedwater mass flow rateFig. 20 Steam generator feedwater temperature3.6 Calculated vs experimental resultsThe main events during the transient are given in Table 2.Table 2 Transient description-sequence of main eventsTime [s]Event-150Start of zerotransient calculation, stabilization of system parameters0Start of transient calculation, E-motor of MCP1 pump switched off3Start of CR group No. 10 movement down from start position of 309.25 cm59Start of CR group No. 9 movement down from start position of 369.25 cm71Reactor power: 1889.04 MW = 62.97 % of Nnom (PSU-REL X-S) 2001.74 MW = 66.01 % of Nnom (KI X-S)72CR group No. 9 in constant position73CR group No. 10 in constant position180Start of CR group No. 9 movement up192CR group No. 9 in upper end position, start of CR group No. 10 movement up193CR group No. 10 in constant position206Start of CR group No. 10 movement up208CR group No. 10 in constant position300End of calculationReactor power: 1940.66 MW = 64.69 % of Nnom (PSU-REL X-S) 2037.38 MW = 67.91 % of Nnom (KI X-S)Comparison of calculated and experimental nuclear powers can be seen in Figure 21. In the case of PSU-REL X-S library the calculated nuclear power coincides practically with the experimental value at the end of transient. The KI X-S library gives a slightly higher power at the final stage of the transient. Fig. 21 Nuclear powerFig. 22 Coolant loop flow rate Fig. 23 Hot leg temperature Fig. 23 Hot leg temperature Comparison of imported calculated and experimental parameters of primary circuit is given Figures 22 through 24. Figures 25 through 32 present the comparison of calculated as well as experimental fuel assembly coolant outlet temperatures. The PSU-REL X-S library gives higher fuel assembly coolant outlet temperatures at the beginning of transient. At the end of transient, we can observe the opposite phenomenon, where fuel assembly coolant outlet temperatures received with KI X-S library are higher. It can be seen in Figures 29 to 32. However, we can say that the calculated temperatures obtained with using of both libraries are in very good agreement especially in the initial phase of transient. Another observed phenomenon is that the calculation results are again, as in Exercise 2, systematically overestimated in comparison with experiment. The maximum deviation is about 10 to 11 oC at the beginning and end of transient. The minimum deviation is approximately 1 oC when the control rods of group No. 9 and 10 are in lower positions. It is shown in Figures 29 through 32. Fig. 25 Fuel assembly coolant outlet temperaturePSU-REL X-SFig. 26 Fuel assembly coolant outlet temperature KI X-SFig. 27 Fuel assembly coolant outlet temperatureKI X-S vs PSU-REL X-SFig. 28 Fuel assembly coolant outlet temperatureKI X-S vs PSU-REL X-S vs experimentFig. 29 Fuel assembly coolant outlet temperaturet = 0 sFig. 30 Fuel assembly coolant outlet temperaturet = 45 sFig. 31 Fuel assembly coolant outlet temperaturet = 71 sFig. 32 Fuel assembly coolant outlet temperaturet = 300 sTemelin NPP – opening of steam dump to atmosphere test exampleThe other experimental transient which can be used for evaluation of fuel assembly coolant outlet temperatures at the NPP with VVER-1000/320 is the steam dump to atmosphere (SDA) test realized at Temelin NPP Unit 2 (Czech Republic). A detailed description of this test and comparison of calculated and experimental results is given in the literature 11 and 12. Only a brief description of the test and used computation tools and neutronic libraries is mentioned here. The main emphasis is placed on the evaluation of fuel assembly coolant outlet temperatures.Transient scenarioTemelin NPP consists from two units VVER-1000/320 type built according original Russian project. The control system was realized by Westinghouse Company (WEC) tools. The fuel assemblies were also of WEC origin at the time of experiment.The experimental test was performed at the beginning of the second fuel cycle of Temelin NPP Unit 2 in April 4, 2004. The reactor power was approximately 20 % of Nnom. All main circulations pumps (MCPs) were in operation. Control rod (CR) groups No. 1 to 9 were in their upper positions. CR group No. 10 was inserted in the core.At the beginning of experimental transient, the main steam header (MSH) pressure was increased with a trend of 1.7 MPa/h (0.028 MPa/min). The increasing or decreasing of MSH pressure was realized with using of steam dump to condenser (SDC). When the steam generator No.1 (SG1) pressure was equal to 6.9 MPa, then the steam dump to atmosphere on the first main steam line (SDA1) was opened. When the SG1 pressure was lower than 6.8 MPa, then the regime of SG1 pressure regulation with constant value of 6.5 MPa (with accuracy of 0.15 MPa) was started. After measurement termination, the final MSH pressure of 6.15 MPa was reached with a trend of -2.5 MPa/h (-0.04 MPa/min). The following thermal-hydraulic phenomena can be observed during the test: SDA opening causes an extensive steam release from affected main steam line with subsequent decreasing of steam temperature and pressure. It leads to decreasing of coolant temperature in attached coolant loop with consequent increasing of reactor power. The unsymmetrical cooling of the reactor causes non-uniform distribution of the power in the core due to the coolant temperature feedback.Computer code usedThe coupled code DYN3D/ATHLET was used for calculation solution of this transient. For the detailed description see Chapters 2.1 and 3.1 as well as references 11 and 12.Reactor core model and nodalization schemesThe reactor core model is fully described by the code DYN3D. It consists of 163 parallel coolant channels where each from them corresponds to one fuel assembly. The reactor core is divided into 27 axial layers (two layers for lower axial reflector, 24 layers for heated part of the core, one layer for upper axial reflector). The zero flux boundary conditions are applied for outer surface of axial reflector. The albedo boundary conditions are specified at radial boundaries of the reactor. Locations of coolant channels in reactor core model and their connection with primary system coolant loops are shown in Figure 33.The standard ATHLET input deck for VVER-1000 Temelin was used. All four loops of reactor coolant system are fully modeled as well as the pressurizer system. The pressurizer surge line is connected to the hot leg of the fourth loop; the spray line is connected with the cold leg of the first loop. The reactor upper plenum was modeled on three levels with four parallel volumes with cross-connections. The space under the reactor pressure vessel head was created by one volume. The lower plenum was modeled as four parallel channels. The core bypass was simulated by the ATHLET code. The secondary circuit modeling corresponds to the primary circuit. There are fully modeled steam generator (SG) secondary side, main steam system (main steam lines (MSL), main steam header (MSH)), and feedwater (FW) system lines.The simplified scheme of reactor vessel, coolant loop No. 1 nodalization and corresponding secondary side is given in Figures 34 to 36.303339532385 DYN3D CORE00 DYN3D CORE Fig. 33 Coolant channels location in the reactor core model Fig. 34 Nodalization of reactor vessel 30581605334000-15213509400 Fig. 35 Coolant loop 1 nodalization Fig. 36 SG1 steam system nodalization4.4 Cross section librariesTwo-group neutronic library ETEWH06 created by ??JV Rez a. s. ? division Energoprojekt Prague was prepared by the help of lattice code HELIOS 13. It contains node-homogenized cross sections and kinetic parameters of fuel assemblies (FA), FA with control clusters, and reflectors. 4.5 Calculated vs experimental resultsResults of simulation of selected global parameters and their comparison with experimental are given in Figures 37 to 40.Fig. 37 Main steam header pressure, calculation vs experimentFig. 38 Steam generator outlet pressure, calculation vs experimentFig. 39 Cold leg temperature, calculation vs experimentFig. 40 Hot leg temperature, calculation vs experimentFigure 41 presents comparison of calculated vs experimental coolant outlet temperatures from different fuel assemblies. The graphs show behavior of outlet temperatures in the reactor core region, which is connected with affected coolant loop No. 1, and in neighboring sectors. Figures 42 through 45 present the comparison of fuel assembly coolant outlet temperatures at the selected time intervals of the transient. The calculation results are systematically lightly overestimated in comparison with experiment. The maximum deviation is about 1.5 to 2 oC mainly at the beginning of transient (see Figure 42). Fig. 41 Fuel assembly coolant outlet temperature, calculation vs experimentFig. 42 Fuel assembly coolant outlet temperature, calculation vs experiment, t = 0 sFig. 43 Fuel assembly coolant outlet temperature, calculation vs experiment, t = 542 sFig. 44 Fuel assembly coolant outlet temperature, calculation vs experiment, t = 1000 sFig. 45 Fuel assembly coolant outlet temperature, calculation vs experiment, t = 1500 sCOnclusionsThe paper presents the calculation of OECD Kalinin-3 benchmark problem with using of stand-alone code DYN3D and externally coupled codes DYN3D/ATHLET. Two different cross-section libraries ? PSU-REL and KI were used. The trends of main calculated parameters (nuclear powers, coolant loops flow rates, coolant loops temperatures) and experimental results are in acceptable agreement ? influence of different X-S libraries can be observed. Differences between calculated fuel assembly coolant outlet temperatures (PSU-REL X-S versus KI X-S library) may be seen. Their ratio is changing during the simulation. Generally we can say that calculated fuel assembly coolant outlet temperatures are higher than experimental values.A question arises. Is the DYN3D and externally coupled codes DYN3D/ATHEL suitable for detailed modeling of coolant outlet distribution in thermocouples location? The clear answer cannot be given in the case of Kalinin-3 benchmark. The thermocouples are located in upper head of fuel assemblies. They are simply modeled as a part of upper axial reflector in the DYN3D model. The same approach is used in the external DYN3D/ATHLET coupling in contrast to detailed modeling used in BIPR/ATHLET coupled codes 14. We can say, that DYN3D and DYN3D/ATHLET fuel assemblies outlet temperatures better reflect the results of thermal balance (t = P/(m c), where P: initial reactor power, m: core mass flow rate, c: mean specific heat, t = 2900000000[W]/163 110[kg/s] 5600 [J/kgoC] 29 oC, tout = tin t = 288 29 = 317 oC, where tout (tin): mean coolant outlet (inlet) temperature at the beginning of transient) than experimental values.The results of SDA test show on a very good agreement of calculated and experimental fuel assembly coolant outlet temperatures. Here we can say that DYN3D/ATHLET external coupling is sufficiently suitable for modeling of this transient. ACKNOWLEDGMENTAuthor would like to thank colleagues from HZDR – Dr. Y. Kozmenkov and Dr. S. Kliem for friendly and valuable consultations during Kalinin-3 Benchmark problem solving. My thanks also to my colleague from JV Rez, a. s. – Dr. Radim Meca for valuable advice in modeling of Temelin NPP SDA test.NOMENCLATURE1-Done-dimensional3-Dthree-dimensionalAERAtomic Energy ResearchATHLETthermal-hydraulic system codeBIPR/ATHLETcoupled code CRcontrol rodFAfuel assemblyDYN3Dreactor dynamic codeGRSGesellschaft fuer Anlagen ?und Reaktor SicherheitHFPhot full powerHPKmain steam headerHSKupper plenumHVShot legHVSKOLhot leg collectorHZDRHelmholtz-Zentrum Dresden-RossendorfKIKurchatov InstituteLPlower plenumLWRlight water reactorMCPmain circulation pumpMSHmain steam headerMSLmain steam lineNnomnominal powerNPP Nuclear Power PlantOECDOrganization for Economic Co-operation and DevelopmentPGsteam generatorPG-PARmain steam linePG-TRsteam generator pipesPSU-RELPenn State University ? Risk Engineering Ltd.PVpressure vesselREFcore bypassRBMKRussian type of reactorSDAsteam dump to atmosphereSDCsteam dump to condenserSGsteam generatorSSRreactor downcomerSVScold legSVSCOLcold leg collectorUPupper plenumVRCHLIKupper headJVstav jadernho vzkumu Rez, a. s. (Nuclear Research Institute Rez, plc.)VVERRussian type of pressurized water reactorWECWestinghouse CompanyX-Scross sectionREFERENCES1V. A. Tereshonok, S. P. Nikonov, M. P. Lizorkin, K. Velkov, A. Pautz, K. Ivanov: Kalinin-3 Coolant Transient Benchmark- Switching-Off of One of the Four Operating Main Circulation Pumps at Nominal Reactor Power, Specification ? First Edition 2008,NEA/NSC/DOC (2009)52 J. Hdek: Solution of the Kalinin-3 Benchmark ? Exercise 2 and 3,24th Symposium of AER on VVER Reactor Physics and Reactor Safety, Sochi, Russian Federation, October 14-18, 20143U. Grundmann, S. Mittag, U. Rohde and S. 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