Simulation of the VVER-1000 pump start-up experiment of the OECD V1000CT benchmark with CATHARE and TRAC-PF1

In the framework of joint effort between the Nuclear Energy Agency (NEA) of OECD, the United States Department of Energy (US DOE), and the Commissariat a l'Energie Atomique (CEA), France a coupled three-dimensional (3D) thermal-hydraulics/neutron kinetics benchmark for VVER-1000 was defined. The benchmark consists of calculation of a pump start-up experiment labelled V1000CT-1 (Phase 1), as well as a vessel mixing experiment and main steam line break (MSLB) transient labelled V1000CT-2 (Phase 2), respectively. The reference nuclear plant is Kozloduy-6 in Bulgaria. The overall objective is to assess computer codes used in the analysis of VVER-1000 reactivity transients. A specific objective is to assess the vessel mixing models used in system codes. Plant data are available for code validation consisting of one experiment of pump start-up (V1000CT-1) and one experiment of steam generator isolation (V1000CT-2). The validated codes can be used to calculate asymmetric MSLB transients involving similar mixing patterns. This paper summarizes a comparison of CATHARE and TRAC-PF1 system code results for V1000CT-1, Exercise 1, which is a full plant point kinetics simulation of a reactor coolant system (RCS) pump start-up experiment. The reference plant data include integral and sector average parameters. The comparison is made from the point of view of vessel mixing and full system simulation. CATHARE used a six-sector multiple 1D vessel thermal-hydraulic model with cross flows and TRAC used a six-sector, 18-channel coarse-mesh 3D vessel model. Good agreement in terms of integral parameters and inter-loop mixing is observed.     

CFD simulation of Fortum PTS experiment

Detailed simulation of the thermal stresses of the reactor pressure vessel (RPV) wall in case of pressurized thermal shock (PTS) requires the simulation of the thermal mixing of cold high-pressure safety injection (HPI) water injected to the cold leg and flowing further to the downcomer. The simulation of the complex mixing phenomena including, e.g., stratification in the cold leg and buoyancy driven plume in the downcomer is a great challenge for CFD methods and requires careful validation of the used modelling methods.The selected experiment of Fortum mixing test facility modelling the Loviisa VVER-440 NPP has been used for the validation of CFD methods for thermal mixing phenomena related to PTS. The experimental data includes local temperature values measured in the cold leg and downcomer. Conclusions have been made on the applicability of used CFD method to thermal mixing simulations in case with stratification in the cold leg and buoyant plume in the downcomer.     

Pre- and post-test calculations to natural circulation experiments at the integral test facility ISB-VVER using the thermalhydraulic code ATHLET

In 1995 at the integral test facility ISB-VVER in Elektrogorsk near Moscow natural circulation experiments were performed, which were scientifically accompanied by the Forschungszentrum Rossendorf. These experiments were the first of this kind at a test facility, which models VVER-1000 thermalhydraulics. Using the code ATHLET which is being developed by ‘Gesellschaft für Anlagen und Reaktorsicherheit’, pre- and post-test calculations were done to determine the thermalhydraulic events to be expected and to define and tune the boundary conditions of the test. The conditions found for natural circulation instabilities and cold leg loop seal clearing could be confirmed by the tests. Besides the thermalhydraulic standard measuring system, the facility was equipped with needle shaped conductivity probes for measuring the local void fractions.     

Fluid mixing and flow distribution in a primary circuit of a nuclear pressurized water reactor—Validation of CFD codes

The EU project FLOMIX-R was 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.This report will focus on the computational fluid dynamics (CFD) code validation. Best practice guidelines (BPG) were applied in all CFD work when choosing computational grid, time step, turbulence models, modelling of internal geometry, boundary conditions, numerical schemes and convergence criteria. The strategy of code validation based on the BPG and a matrix of CFD code validation calculations have been elaborated. CFD calculations have been accomplished for selected experiments with two different CFD codes (CFX, FLUENT). The matrix of benchmark cases contains slug mixing tests simulating the start-up of the first main circulation pump which have been performed with three 1:5 scaled facilities: the Rossendorf coolant mixing model ROCOM, the Vattenfall test facility and a metal mock-up of a VVER-1000 type reactor. Before studying mixing in transients, ROCOM test cases with steady-state flow conditions were considered. Considering buoyancy driven mixing, experimental results on mixing of fluids with density differences obtained at ROCOM and the FORTUM PTS test facility were compared with calculations. Methods for a quantitative comparison between the calculated and measured mixing scalar distributions have been elaborated and applied. Based on the “best practice CFD solutions”, conclusions on the applicability of CFD for turbulent mixing problems in PWR were drawn and recommendations on CFD modelling were given. The results of the CFD calculations are mostly in-between the uncertainty bands of the experiments. Although no fully grid-independent numerical solutions could be obtained, it can be concluded about the suitability of applying CFD methods in engineering applications for turbulent mixing in nuclear reactors.     

Modeling of a buoyancy-driven flow experiment at the ROCOM test facility using the CFD codes CFX-5 and Trio_U

The influence of density differences on the mixing of the primary loop inventory and the emergency core cooling (ECC) water in the downcomer of a pressurized water reactor (PWR) was analyzed at the ROssendorf COolant Mixing (ROCOM) test facility. ROCOM is a 1:5 scaled model of a German PWR, and has been designed for coolant mixing studies. It is equipped with advanced instrumentation, which delivers high-resolution information for temperature or boron concentration fields.An experiment with 5% of the design flow rate in one loop and 10% density difference between the ECC and loop water was selected for validation of the CFD software packages CFX-5 and Trio_U. Two similar meshes with approximately 2 million control volumes were used for the calculations. The effects of turbulence on the mean flow were modeled with a Reynolds stress turbulence model in CFX-5 and a LES approach in Trio_U. CFX-5 is a commercial code package offered from ANSYS Inc. and Trio_U is a CFD tool which is developed by the CEA-Grenoble, France.The results of the experiment and of the numerical calculations show that mixing is dominated by buoyancy effects: at higher mass flow rates (close to nominal conditions) the injected slug propagates in the circumferential direction around the core barrel. Buoyancy effects reduce this propagation. The ECC water falls in an almost vertical path and reaches the lower downcomer sensor directly below the inlet nozzle. Therefore, density effects play an important role during natural convection with ECC injection in PWRs. Both CFD codes were able to predict well the observed flow patterns and mixing phenomena.     

Detailed CFX-5 study of the coolant mixing within the reactor pressure vessel of a VVER-1000 reactor during a non-symmetrical heat-up test

As part of the reactor dynamics activities of FZK/IRS, the qualification of a detailed 3D CFD model of a reactor pressure vessel is a key step in safety evaluations for improving predictive capabilities and acceptability of commercial CFD tools in reactor physics. The VVER-1000 Coolant Transient Benchmark, initiated by OECD, represents an excellent opportunity for validation. In this work a CFD model for the complete VVER-1000 reactor pressure vessel is presented. Due to computational limits simplifications of the core and of some other geometrical details are introduced. The simulated scenario is the heat-up of one coolant loop in case of the isolation of a steam generator while the reactor is operating at a low power level. Two transient runs with a first and second order approximation of the spatial discretization are performed. Unexpectedly, the first order method reveals better agreement with measured data.     

Analysis of natural circulation phenomena in VVER-1000

In all light water reactors (LWR), natural circulation is an important passive heat removal mechanism. In the present paper, the natural circulation phenomena are studied with reference to step-wise coolant inventory reduction and a small break loss-of-coolant-accident (SBLOCA) in the cold leg of VVER-1000. The natural circulation flow map (NCFM) approach is considered to evaluate the natural circulation performance of the VVER-1000 NPP also comparing VVER-1000 and PWR systems. Three different elevations between heat source (core) and heat sink (steam generators) zones have been considered in order to characterize the buoyancy force in a VVER-1000. The influence of power and the cold legs loop seal upon the natural circulation performance is also evaluated. In the second part, a series of SBLOCA simulations with break area ranging from 0.5 to 11.7% of the cold leg cross sectional area are performed starting with the VVER-1000 system in nominal conditions. The effect of Emergency Core Cooling System (ECCS) including passive and active parts of ECCS are evaluated. The simulations were performed by the help of the system code RELAP5. Within the framework of the qualification of the adopted computational tools, the results are compared with experimental data from Kozloduy NPP unit 6 test and PSB-VVER integral test facility available from the literature. Namely, the qualification of the adopted nodalisation in steady state conditions is achieved by using experimental data. The accuracy of selected results have been estimated in quantitative terms by applying the fast Fourier transform based method (FFTBM). Finally, the relevance and the potential for the occurrence of the reflux condensation mode, i.e., one of the Natural Circulation regimes, for cooling of reactor core in VVER-1000 are discussed.     

Natural circulation flow behavior at reduced inventory in a VVER geometry

This paper deals with the natural circulation flow characteristics of the VVER-440 geometry at reduced coolant inventory. Special emphasis is on the flow rate of the primary circuits during the two-phase flow regime. For studying two-phase natural circulation flow phenomena in a VVER geometry a series of cold leg small break loss-of-coolant accident (SBLOCA) tests was carried out in the PArallel Channel TEst Loop (PACTEL), a 1/305 volumetrically scaled model of a VVER-440 reactor. The tests were conducted with break areas ranging from 0.1 to 1.5 % of the scaled cold leg cross-sectional area of the reference reactor. A partial failure of the high-pressure injection system (HPIS) was assumed. The tests reveal a trend towards an increasing primary circuit mass flow rate with decreasing inventory. This contradicts the findings of earlier tests in multi-loop VVER geometry. With single-loop facilities, increased mass flow rates at reduced inventories have been reported before. The increase of the two-phase flow rate turns out to be a consequence of the combined effect of break size, pressure range and secondary side feed and bleed procedure. The physical phenomena of flow stagnation in the primary circuits, system pressurization, asymmetric loop flows, and loop seal clearing and refilling take place during the natural circulation cooling process from single-phase into two-phase and boiler–condenser modes. In addition, flow reversal in the undermost tubes of the horizontal steam generators (SG) is observed. These phenomena are discussed briefly while a general insight into the course of the tests is presented.     

Boron dilution transients during natural circulation flow in PWR—Experiments and CFD simulations

Coolant mixing in the cold leg, downcomer and the lower plenum of pressurized water reactors is an important phenomenon mitigating the reactivity insertion into the core. Therefore, mixing of the de-borated slugs with the ambient coolant in the reactor pressure vessel was investigated at the four loops 1:5 scaled Rossendorf coolant mixing model (ROCOM) mixing test facility. In particular thermal hydraulics analyses have shown, that weakly borated condensate can accumulate in the pump loop seal of those loops, which do not receive a safety injection. After refilling of the primary circuit, natural circulation in the stagnant loops can re-establish simultaneously and the de-borated slugs are shifted towards the reactor pressure vessel (RPV).In the ROCOM experiments, the length of the flow ramp and the initial density difference between the slugs and the ambient coolant was varied. From the test matrix experiments with 0 resp. 2% density difference between the de-borated slugs and the ambient coolant were used to validate the CFD software ANSYS CFX. To model the effects of turbulence on the mean flow a higher order Reynolds stress turbulence model was employed and a mesh consisting of 6.4 million hybrid elements was utilized. Only the experiments and CFD calculations with modeled density differences show stratification in the downcomer. Depending on the degree of density differences the less dense slugs flow around the core barrel at the top of the downcomer. At the opposite side, the lower borated coolant is entrained by the colder safety injection water and transported to the core. The validation proves that ANSYS CFX is able to simulate appropriately the flow field and mixing effects of coolant with different densities.     

Experiments at the mixing test facility ROCOM for benchmarking of CFD codes

For the validation of computational fluid dynamics (CFD) codes, experimental data on fluid flow parameters with high resolution in time and space are needed.Rossendorf Coolant Mixing Model (ROCOM) is a test facility for the investigation of coolant mixing in the primary circuit of pressurized water reactors. This facility reproduces the primary circuit of a German KONVOI-type reactor. All important details of the reactor pressure vessel are modelled at a linear scale of 1:5. The facility is characterized by flexible possibilities of operation in a wide variety of flow regimes and boundary conditions. The flow path of the coolant from the cold legs through the downcomer until the inlet into the core is equipped with high-resolution detectors, in particular, wire mesh sensors in the downcomer of the vessel with a mesh of 64 × 32 measurement positions and in the core inlet plane with one measurement position for the entry into each fuel assembly, to enable high-level CFD code validation. Two different types of experiments at the ROCOM test facility have been proposed for this purpose. The first proposal concerns the transport of a slug of hot, under-borated condensate, which has formed in the cold leg after a small break LOCA, towards the reactor core under natural circulation. The propagation of the emergency core cooling water in the test facility under natural circulation or even stagnant flow conditions should be investigated in the second type of experiment. The measured data can contribute significantly to the validation of CFD codes for complex mixing processes with high relevance for nuclear safety.     

Comparison of measured and predicted thermal mixing tests using improved finite difference technique

Experimental and computational analyses of a mixing test of cold and hot water flows in a rectangular tee model of the cold leg downcomer geometry of pressurized water reactor were performed. Results obtained from COMMIX-1A computer code calculations showed reasonable agreement with the experimental findings. Counter-current flow and thermal stratification in the cold leg were observed in both the experimental and calculated results for certain ranges of test parameters.     

Experimental and computational thermal-hydraulic results of fluid and thermal mixing tests

Experimental and computational analyses of a mixing test of cold and hot water flows in a rectangular tee model of the cold leg downcomer geometry of pressurized water reactor were performed. Results obtained from COMMIX-1A computer code calculations showed reasonable agreement with the experimental findings. Counter-current flow and thermal stratification in the cold leg were observed in both the experimental and calculated results for certain ranges of test parameters.     

Primary loop study of a VVER-1000 reactor with special focus on coolant mixing

The intention of this work is to demonstrate the ability of modern computational fluid dynamics (CFD) tools like ANSYS CFX to model technical facilities with complex geometry such as a reactor pressure vessel with its primary loops over all relevant scales from a range of approximately 1 mm up to a largest dimension of about 50 m. For this purpose, a detailed model of a VVER1000 reactor pressure vessel (RPV) and an extended model with simplified primary loops (Loop model) are presented. The primary loop components like steam generators and pumps are modelled by the outer shapes and additional source terms for energy and momentum exchange. The RPV model part is tested without primary loops; results are compared with former results obtained with a coarser model. The improved model shows less sensitivity on the discretization scheme, and recalculated hot leg temperatures transients are improved. As an example case for the Loop model, the swirl of flow patterns at the core inlet derived from experimental data obtained at the Kozloduy Nuclear Power Plant (Unit 6) is simulated. As progress to the RPV model the Loop model is able to predict a swirl, but due to the lack of technical details, like the pump impellers, the swirl is overestimated.     

Analyses of the V1000CT-1 benchmark with the DYN3D/ATHLET and DYN3D/RELAP coupled code systems including a coolant mixing model validated against CFD calculations

Plant-measured data provided within the specification of the OECD/NEA VVER-1000 coolant transient benchmark (V1000CT) were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to the MCP (main coolant pump) switching on experiment conducted in the frame of the plant-commissioning activities at the Kozloduy NPP Unit 6 in Bulgaria. The experiment was started at the beginning of cycle (BOC) with average core expose of 30.7 effective full power days (EFPD), when the reactor power was at 27.5% of the nominal level and three out of four MCPs were operating. The transient is characterized by a rapid increase in the primary coolant flow through the core and, as a consequence, a decrease of the space-dependent core inlet temperature. Both DYN3D/RELAP5 and DYN3D/ATHLET analyses were based on the same reactor model, including identical MCP characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. For an adequate modelling of the redistribution of the coolant flow in the reactor pressure vessel during the transient a simplified mixing model for the DYN3D/ATHLET code was developed and validated against a computational fluid dynamics calculation.

The results of both coupled code calculations are in good agreement with the available experimental data. The discrepancies between experimental data and the results of both coupled code calculations do not exceed the accuracy of the measurement data. This concerns the initial steady-state data as well as the time histories during the transient. In addition to the validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has been performed to evaluate relevant thermal hydraulic models of the system codes RELAP5 and ATHLET and to explain differences between the calculation results.     

Simulation of the VVER-1000 pump start-up experiment in the OECD/DOE/CEA V1000CT benchmark by the FLICA4/CRONOS2 coupled code system

In the framework of joint effort between the Nuclear Energy Agency (NEA) of OECD, the United States Department of Energy (US DOE), and the Commissariat à l'Energie Atomique (CEA), France, a coupled 3-D thermal–hydraulics/neutron kinetics benchmark was defined. The overall objective of OECD/NEA V1000CT benchmark [Ivanov, B., Ivanov, K., Groudev, P., Pavlova, M., Hadjiev, V., 2002. VVER-1000 Coolant Transient Benchmark (V1000-CT). Phase 1 – Final Specifications, NEA/NSC/DOC] is to assess computer codes used in the analysis of VVER-1000 reactivity transients where mixing phenomena (mass flow and temperature) in the reactor pressure vessel are complex. Original data from the Kozloduy-6 Nuclear Power Plant are available for the validation of computer codes: one experiment of pump start-up (V1000CT-1) and one experiment of steam generator isolation (V1000CT-2). The CEA presented results for the V1000CT-1 Exercise 2 using a coupling of FLICA4 [Toumi, I., Gallo, D., Bergeron, A., Royer, E., Caruge, D., 2000. FLICA4: a three dimensional two-phase flow computer code with advanced numerical methods for nuclear applications. Nuclear Engineering and Design 200, 139–155] and CRONOS2 [Akherraz, B., Baudron, A.M., Lautard, J.J., Magnaud, C., Moreau, F., Schneider, D., Gonzales, M., 2004. Manuel de Référence CRONOS 2.6. Technical Report SERMA/LENR/RT/04-3433/A, CEA] via the coupling tool ISAS [Toumi, I., et al., 1995. Specifications of the general software architecture for code integration in ISAS. Euratom Fusion Technology, ITER task S81TT-01/1]. The FLICA4/CRONOS2 VVER-1000 model is based on the data available in the benchmark specifications. This paper summarizes the FLICA4/CRONOS2 model build-up with the associated sensitivity studies and presents the CEA results for V1000CT-1 Exercise 2 as well as a comparison with experimental results at hot power steady state (HP SS).     

RELAP5/MOD3.2 investigation of reactor vessel YR line capabilities for primary side depressurization during the TLFW in VVER1000/V320

During the development of Symptom Based Emergency Operating Procedures (SB-EOPs) for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (NPP), a number of analyses have been performed using the RELAP5/MOD3.2 computer code. One of them is “Investigation of reactor vessel YR line capabilities for primary side depressurization during the Total Loss of Feed Water (TLFW)”. The main purpose of these calculations is to evaluate the capabilities of YR line located at the top of the reactor vessel for primary side depressurization to the set point of High Pressure Injection System (HPIS) actuation and the abilities for successful core cooling after Feed&Bleed procedure initiation. For the purpose of this, operator action with “Reactor vessel off-gas valve – 0.032 m” opening has been investigated. RELAP5/MOD3.2 computer code has been used to simulate the TLFW transient in VVER-1000 NPP model. This model was developed at Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events, and design basis scenarios. The model provides a significant analytical capability for the specialists working in the field of NPP safety.     

Numerical boron mixing studies for Loviisa nuclear power plant

A program is in the process of studying numerically boron mixing in the downcomer of Loviisa NPP (VVER-440). Mixing during the transport of a diluted slug from the loop to the core might serve as an inherent protection mechanism against severe reactivity accidents in inhomogenous boron dilution scenarios for PWRs. The commercial general purpose Computational Fluid Dynamics (CFD) code PHOENICS is used for solving the governing fluid flow equations in the downcomer geometry of VVER-440. So far numerical analyses have been performed for steady state operation conditions and two different pump driven transients. The steady state analyses focused on model development and validation against existing experimental data. The two pump driven transient scenarios reported are based on slug transport during the start of the sixth and first loop, respectively. The results from the two transients show that mixing is case and plant specific; the high and open downcomer geometry of VVER-440 seems to be advantageous from mixing point of view. In addition the analyzing work for the ‘first pump start' scenario brought up some considerations about flow distribution in the existing experimental facilities.     

AFEN method and its solutions of the hexagonal three-dimensional VVER-1000 benchmark problem

The unique features of the analytic function expansion nodal (AFEN) method in hexagonal-z geometry are described. The COREDAX code implementing the AFEN method is verified testing on the VVER-440 benchmark problem and a “simplified” VVER-1000 benchmark problem. The COREDAX code then applied to the original VVER-1000 benchmark problem exercise 2 (HZP case and HP case) provides very good results in comparison with those of other benchmark participants.