International Conference Nuclear Energy for New Europe 2005 Bled, Slovenia, September 5-8, 2005 VVER-1000 Coolant Transient Benchmark - Overview and Status of Phase 2 Nikola Kolev, Nikolay Petrov Institute for Nuclear Research and Nuclear Energy 72 Tzarigradsko chaussee, Blvd., BG-1784 Sofia, Bulgaria npkolev@inrne.bas.bg, nlp@ mail.bg Sylvie Aniel, Eric Royer Commissariat à l Energie Atomique DM2S/SFME, Centre de Scalay F-91191 Gif sur Yvette Cedex, France sylvie.aniel@cea.fr, eric.royer@cea.fr Ulrich Bieder Commissariat à l Energie Atomique DER/SSTH, Centre de Grenoble 17 rue des Martyrs, F-38054 Grenoble Cedex 9, France ulrich.bieder@cea.fr ABSTRACT The Main Steam Line Break (MSLB) is identified as a Design Basis Accident (DBA) exhibiting significant localized space-time effects. There is a need to validate the models used in thermal-hydraulics and neutronics, and particularly the flow mixing in the core vessel. For this purpose, a consistent approach is defined with three exercises. First, a Coolant Mixing test carried out at the Kozloduy Nuclear Power Plant (KNPP) Unit 6 is used to validate the thermal-hydraulics of the vessel. Several types of models are possible, from Large Eddy Scale to multi-channels. Then two scenarios of MSLB are specified to check the coupling between core vessel thermal-hydraulics and neutronics on one hand, and between vessel and plant on the other hand. The paper presents an overview of the V1000CT-2 benchmark exercises and currently available results. 1 INTRODUCTION In the framework of joint effort between the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD), the United States Department of Energy (US DOE), and the Commissariat à l Energie Atomique (CEA), France a coupled 3-D neutron kinetics/thermal hydraulics benchmark was defined [2]. The benchmark is based on data from the Unit 6 of the Bulgarian Kozloduy NPP. In performing this work the Pennsylvania State University (PSU), USA and CEA-Saclay, France have collaborated with Bulgarian organizations, in particular with KNPP and the Institute for 116.1
116.2 Nuclear Research and Nuclear Energy (INRNE). The benchmark consists of two phases: Phase 1: Main Coolant Pump Switching On; Phase 2: Coolant Mixing Tests and MSLB. Since the previous coupled code benchmarks [1] indicated that further development of the mixing computation models in the integrated codes is necessary, a coolant mixing experiment and MSLB transients were selected for simulation in Phase 2 of the benchmark. The MSLB event is characterized by a large asymmetric cooling of the core, stuck rods and a large primary coolant flow variation. Two scenarios are defined: the first scenario is taken from the current licensing practice and the second is derived from the original one using aggravating assumptions to enhance the code-to-code comparisons. Three workshops have taken place since the starter benchmark workshop in Dresden. The first workshop was held in Saclay, France in May 2003, the second workshop was in Sofia, Bulgaria in April 2004, and the latest one in Garching, Germany in April 2005. During the third workshop, the final results for V1000CT-1 were discussed, and preliminary results for V1000CT-2 Exercise 1 were presented. This paper aims at presenting the three exercises of the benchmark, discuss the modelling issues of temperature mixing in the core vessel, and the application to a MSLB analysis. 2 DESCRIPTION OF V1000CT-2 EXERCISES 2.1 Exercise 1: Computation of coolant mixing experiments This exercise is based on a comparison with a mixing experiment conducted at Kozloduy-6 as part of the plant-commissioning phase. The experiment includes isolation of a steam generator at 9.3% of the nominal power causing single loop heat-up, with all MCP in operation. It is characterized by temperature rise of about 14 degrees and a decrease of mass flow rate by 3.4% in the disturbed loop, affecting the neighbouring loops as well. It will be used to test and validate vessel-mixing models (CFD, coarse-mesh and mixing matrix). Vessel boundary conditions and core power distribution along with pressure above the core will be part of the exercise specification. A particular effort was necessary to define the vessel geometry, because plant specific data appeared to be sensitive on the flow mixing. The task is to calculate the core inlet and outlet distributions. 2.2 Exercises 2 and 3: Main Steam-Line Break (MSLB) modelling The transient to be analyzed is initiated by a main steam line break in a VVER-1000 between the steam generator (SG) and the steam isolation valve (SIV), outside the containment. A mechanical failure of the main feed water regulation valve is assumed. This event is characterized by a large asymmetric cooling of the core, stuck control rods and a large primary coolant flow variation. Two scenarios will be defined: the first scenario is taken from the current licensing practice and the second is derived from the original one using aggravating assumptions to enhance the code-to-code comparison. The main objective of the study is to clarify the local 3-D feedback effects depending on the vessel mixing. Special emphasis is put on testing 3-D vessel thermal-hydraulic (T-H) models and the coupling of 3-D neutronics/vessel thermal hydraulics. The MSLB is thus divided in two exercises (to be done for the two scenarios): Exercise 2 consists of coupled 3-D neutronics/vessel thermal-hydraulic simulations using specified vessel T-H boundary conditions and Exercise 3 consists of best estimate coupled plant simulations (plant, 3-D vessel and core).
116.3 3 STATUS AND PRELIMINARY RESULTS 3.1 Exercise 1 Specifications for Exercise 1 have been delivered in March, 2004, after a comprehensive verification of the geometrical data of the pressure vessel of KNPP. These data are available both in spreadsheet format with tables and drawings, and as a CAD geometry file. The geometry of both, the reactor vessel and of the lower plenum with the 163 support columns is presented in Figure 1. Cold leg nozzle Zoom of the Lower Plenum Downcomer Consols Support Column Core Plate Elliptic RPV Narrowing Gap Support Column Perforated Barrel Figure 1: Geometry of the VVER-1000 reactor There are currently six participants who have already obtained preliminary results, and in addition six other participants (in italic in Table 1) who have intention to perform the calculations. Table 1: List of participants for V1000CT-2 Exercise 1 Company Country Code CEA France Trio_U FZR Germany CFX 5 FZK Germany CFX 5 Kurchatov Institut Russia ATHLET Penn State and ORNL USA RELAP 3D INRNE Bulgaria CATHARE GRS Germany ATHLET University of Pisa Italy RELAP VTT Finland Porflo NRI Czech Republic Fluent Technical University of Budapest Hungary CFX 5 EREC, Electrogorsk Russia In-house CFD code
116.4 Experimental data show a net counter clock-wise shift (rotation) of "loop flow centres" in certain VVER-1000 V320 flow patterns. The swirl intensity is depending on the considered unit. The origin of the swirl is not clearly identified, however the asymmetry of the actual vessel and internals due to the fabrication process as well as local flow disturbances are expected to be the main sources of the swirl. CFD computations with the Trio_U code [4] have been carried out by CEA [3]. The experimentally detected swirl can be reproduced with the Trio_U code by using a tetrahedral mesh of about 10.000.000 control volumes and an under-resolved LES turbulence modelling approach. The calculated and measured temperature distribution at the core inlet is given in Figure 2. Here, the stabilized thermal hydraulic situation about 20 minutes after the isolation of SG-1 is shown. The locations of the cold legs are also added to this Figure. The calculated asymmetric temperature field and the counter clockwise rotation of 24 of the temperature maximums with respect to the axis of cold leg 1 are in excellent agreement with the experimental data. Experiment Trio_U Calculation Leg 1 Leg 1 Figure 2 : Temperature distribution at the core inlet at the end of the test The experimental values include the mixing in the core region, what is not taken into account in the calculation. This explains the enhanced mixing in the experiment at the periphery of the zone affected by leg 1.
116.5 284 282 280 Trio_U Experiment Temperature 278 276 274 272 270 268 1 21 41 61 81 101 121 141 161 Assembly Figure 3: Assembly by assembly comparison of the fluid temperature ( C) A more quantitative evaluation is given in Figure 3. The measured and calculated mean temperature of the 163 assemblies are compared, where the numbering goes line by line from the lower left to the upper right assembly (see Figure 2). Exercises 2 and 3 MSLB situations have been computed by INRNE with the CATHARE 1.3 code [5], in order to specify the scenarios. The core vessel model is based on four channels, and has been validated against the pump start-up problem (Phase 1 of the benchmark) and the coolant mixing problem (Phase 2, Exercise 1). The results have been discussed during the third workshop in Garching. The initial conditions for the accident correspond to the end of Cycle 8 at nominal power. The hypotheses for the realistic scenario are: a double ended break on Steam Line 4, upstream of the isolation valve, the SG water inventory is maximal, the SG-4 feedwater regulating valve is stuck in fully open position, the most effective control rod is not inserted when scram occurs, and boron of HPSI is not taken into account. This scenario, labelled Scenario 1, is very close to what is used in the current licensing practice. A preliminary calculation of the reactor vessel boundary conditions is shown in Figures 4 through 6. There is no return to power, because the Main Coolant Pump 4 trips, resulting in a reverse flow in loop 4 (Figure 5) and a limited cooling of the core (Figure 4). In order to enhance the reactivity insertion (return to power after scram, Figure 7) and the multi-dimensional effects, two extra hypotheses are considered for the pessimistic scenario labelled Scenario 2: Main Coolant Pump 4 does not trip on signal, and a second control rod remains stuck out of the core. Preliminary results for the reactor power history and vessel boundary conditions are shown in Figures 7 through 10. The temperature at cold leg 4 drops to 210 C (Figure 8).The preliminary calculations are used to derive the vessel boundary
116.6 conditions for Exercise 2, and also to define the range of parameters for the cross-section libraries: fuel temperature, moderator density and temperature. Cross-section libraries for the reference core are being prepared with the Helios 1.7 lattice code in PSU-INRNE-KNPP collaboration. Figure 4: History of cold leg temperatures for Scenario 1 (realistic) Figure 5: History of vessel inlet flows for Scenario 1 (realistic)
116.7 Figure 6: History of vessel outlet pressures for Scenario 1 (realistic) Figure 7: History of relative power for Scenario 2 (pessimistic)
116.8 Figure 8: History of cold leg temperatures for Scenario 2 (pessimistic) Figure 9: History of vessel inlet flows for Scenario 2 (pessimistic)
116.9 Figure 10: History of vessel outlet pressure for Scenario 2 (pessimistic) 4 CONCLUSIONS AND PERSPECTIVES V1000CT-2 benchmark consists of three exercises. The first exercise, dedicated to the computation of a flow mixing experiment in the core vessel, is under way for the participants. Several types of thermal-hydraulic models are used, ranging from Large Eddy Scale to multi- 1D channel. The final results are expected by September 2005, in order to be analysed by the benchmark team, and finally presented during the fourth workshop in April 2006. The second and third exercises are devoted to the simulation of a Main Steam Line Break accident. Two scenarios are currently under definition for the participants: the realistic one is close to the licensing practice, whereas the pessimistic one assumes aggravating events: the Main Coolant Pump of the faulted loop does not trip on signal and a second control rod remains stuck-out of the core when the scram occurs. The pessimistic scenario results in a significant return to power and well pronounced multi-dimensional effects in the core. Participants results for exercises 2 and 3 of V1000CT-2 are expected in 2006 in order to have a comparative analysis. The conclusions will be reported in OECD documents. ACKNOWLEDGMENTS The authors thank all the members of the benchmark team for their support. REFERENCES [1] Ivanov, K., Beam, T., Baratta, A., Irani, A., and Trikorous, N., PWR MSLB Benchmark. Volume 1: Final Specifications, NEA/NSC/DOC (99) 8 [2] Ivanov, B., Ivanov, K., Royer, E., Aniel, U., Kolev, N. and Groudev, P. OECD/DOE/CEA VVER-1000 coolant transient (V1000CT) benchmark for assessing coupled neutronics/thermal-hydraulics system codes for VVER-1000 RIA analysis, Proc. PHYSOR 2004, April 25-29, 2004, Chicago, IL USA
116.10 [3] Bieder, U., Fauchet, G., Betin, S., Kolev, N., Popov, D., Simulation of mixing effects in a VVER-1000 reactor, Proc. NURETH 11 conference, October 2-6, 2005, Avignon, France [4] Bieder, U., Calvin C., Mutelle H., Detailed thermal-hydraulic analysis of induced break severe accidents using the massively parallel CFD code Trio_U. Int. Conf. on Supercomputing in Nuclear Applications, Paris 22-24 September 2003 [5] Bestion, D., The physical closure laws in the CATHARE code, Nuclear Engineering and Design, vol. 124, Dec. 1990, pp. 229-245