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.     

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.     

Experimental determination of the boron concentration distribution in the primary circuit of a PWR after a postulated cold leg small break loss-of-coolant-accident with cold leg safety injection

It is known that under-borated coolant can accumulate in the loops and that it can be transported towards the reactor core during a loss-of-coolant-accident. Therefore, the mixing of weakly borated water inside the reactor pressure vessel was investigated using the ROCOM test facility. Wire-mesh sensors based on electrical conductivity measurement are used to measure in detail the spreading of a tracer solution in the facility. The mixing in the downcomer was observed with a measuring grid of 64 azimuthal and 32 vertical positions. The resulting distribution of the boron concentration at the core inlet was measured with a sensor integrated into the lower core support plate providing one measurement position at the entry into each fuel assembly.

The boundary conditions for this mixing experiment are taken from an experiment at the thermal hydraulic test facility PKL operated by AREVA Germany. The slugs, which have a lower density, accumulate in the upper part of the downcomer after entering the vessel. The ECC water injected into the reactor pressure vessel falls almost straight down through this weakly borated water layer and accelerates as it drops over the height of the downcomer. On the outer sides of the ECC streak, lower borated coolant admixes and flows together with the ECC water downwards. This has been found to be the only mechanism of transporting the lower borated water into the lower plenum. In the core inlet plane, a reduced boron concentration is detected only in the outer reaches of the core inlet. The minimum instantaneous boron concentration that was measured at a single fuel element inlet was found to be 66.3% of the initial 2500 ppm.     

High-resolution boron dilution measurements using laser induced fluorescence (LIF)

The application of the laser induced fluorescence technique to the study of liquid mixing in the downcomer of a pressurized water reactor is presented. The scenario is that of a boron dilution event, in which a deborated slug is set in motion by the actuation of a reactor coolant pump. A separate effects test facility, built with transparent plexiglas, is used to conduct optical measurements of the slug mixing along its path to the core. The optical assembly is described and the conditions for the implementation of laser induced fluorescence as a quantitative measurement technique are discussed. Results from a slug injection experiment are shown which demonstrate the high-resolution capabilities of this procedure as applied to the study of liquid mixing in the complex geometry of a reactor vessel downcomer.     

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.     

An integral loop test and MARS code analysis for a DVI line break LOCA in the APR1400

The transient thermal-hydraulic phenomena of a DVI (Direct Vessel Injection) line break LOCA (Loss-of-Coolant Accident) in pressurized water reactor, APR1400, were investigated. In order to understand the phenomena during the LOCA transient, a reduced-height and reduced-pressure integral loop test facility, the SNUF (Seoul National University Facility), was constructed with scaling down the prototype. For the appropriate test conditions in the experiment with the SNUF, the energy scaling method was suggested with scaling the coolant mass inventory and the thermal power for the reduced-pressure condition. According to the conditions determined by the method, the experimental study was performed with the SNUF. The experimental results showed that the phenomenon of the downcomer seal clearing played a dominant role in the reduction of the system pressure and the recovery of the coolant level in the core. That phenomenon occurred when the steam incoming from cold legs penetrates the coolant in the upper downcomer toward the broken DVI line. The experimental results were compared with the prototype analysis to estimate the energy scaling method, so that the experiment reasonably simulated the phenomena in the prototype. For the analytical investigation, the experiment was simulated with MARS code to validate the calculation capability of the code, especially for the downcomer seal clearing, which showed good agreement with the results of experiment.     

Numerical analysis and visualization experiment on behavior of borated water during MSLB with RCP running mode in an advanced reactor

The core bypass phenomenon of borated water injected through direct vessel injection (DVI) nozzles in APR1400 (Advanced Power Reactor 1400MWe) during main steam line break (MSLB) accidents with a reactor coolant pump (RCP) running mode has been simulated using a two-channel and one-dimensional system analysis model code (MARS), and a three-dimensional computational fluid dynamics (CFD) code (FLUENT). A visualization experiment has also been performed using a scaled-down model of the APR1400. The MARS analysis has predicted a serious core bypass phenomenon of borated water, while the CFD analysis has shown results opposite to the MARS results. The CFD analysis has shown that the flow pattern in the downcomer is fully three-dimensional and that vortex flow structures are formed near the cold legs so that the borated water might pass without difficulty into the high flow region of the cold legs and flow well into the lower downcomer. The visualization experiment has shown that the borated water flows well to the lower plenum, as in the CFD analysis. Both the CFD analysis and visualization experiment have proved that a serious core bypass phenomenon of borated water might not happen in the APR1400. These results are quite different from those predicted by MARS.     

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.     

Simulation of mixing effects in a VVER-1000 reactor

This work has been performed in the framework of the OECD/NEA thermalhydraulic benchmark V1000CT-2. This benchmark is related to fluid mixing in the reactor vessel during a MSLB accident scenario in a VVER-1000 reactor. Coolant mixing in a VVER-1000 V320 reactor was investigated in plant experiments during the commissioning of the Unit 6 of the Kozloduy nuclear power plant. Non-uniform and asymmetric loop flow mixing in the reactor vessel has been observed in the event of symmetric main coolant pump operation. For certain flow conditions, the experimental evidence of an azimuthal shift of the main loop flows with respect to the cold leg axes (swirl) was found.Such asymmetric flow distribution was analyzed with the Trio_U code. Trio_U is a CFD code developed by the CEA Grenoble, aimed to supply an efficient computational tool to simulate transient thermalhydraulic turbulent flows encountered in nuclear systems. For the presented study, a LES approach was used to simulate turbulent mixing. Therefore, a very precise tetrahedral mesh with more than 10 million control volumes has been created.The Trio_U calculation has correctly reproduced the measured rotation of the flow when the CAD data of the constructed reactor pressure vessel where used. This is also true for the comparison of cold leg to assembly mixing coefficients. Using the design data, the calculated swirl was significantly underestimated. Due to this result, it might be possible to improve with CFD calculations the lower plenum flow mixing matrices which are usually used in system codes.     

CFD study on coolant mixing in VVER-440 fuel rod bundles and fuel assembly heads

A CFD model of VVER-440 fuel assembly heads was developed based on the technical documentation of a full-scale test facility built in the Kurchatov Institute, Russia. Steady-state and transient calculations were performed to validate the model with a measurement set. Effects of the spatial resolution, turbulence models, difference schemes and different inlet boundary conditions were investigated. Inlet boundary conditions were determined with both the COBRA subchannel code and a fuel rod bundle CFD model that was built for this special purpose. The results were compared against experimental data. The sensitivity studies showed that a grid of about 8 million cells, high resolution scheme and BSL Reynolds stress model are suitable sets to provide accurate prediction for the signal of the in-core thermocouple. The best prediction was achieved with transient calculation using inlet boundary conditions generated with the CFD fuel rod bundle model. The results indicated that the coolant mixing is intensive but not perfect in the assembly head. Besides, the significant role of the outflow from the central tube was also proven. The transient runs revealed relatively large temperature fluctuations near the in-core thermocouple housing.     

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.     

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.     

高压安注条件下冷管段和环腔流体混合特性的数值分析

压水堆高压安注条件下冷热流体混合会导致承压热冲击现象,影响压力容器的使用寿命。本文基于ROCOM实验装置的实验数据,使用CFD方法对高压安注条件下有密度差的冷热流体混合现象进行了模拟,并对模拟结果进行了验证与分析。结果表明,在冷管段和下降段环腔中流体混合的主导因素分别为强迫流动混合和浮升力驱动混合。在仅有1条冷管段注入的情况下,进入下腔室的流体会再次回流至环腔,从而对冷却剂的混合特性产生影响。     

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.     

Experimental studies and CFD calculations for buoyancy driven mixing phenomena

In nuclear reactor safety the mixing of borated and deborated water is a critical issue that needs investigation, assessment and prediction. Such mixing is buoyancy driven and numerical codes must correctly model momentum transfer between fluids of different density. To assess and develop CFD models for buoyancy driven mixing we set up a simple vertical mixing test facility (VeMix) and equipped it with a newly developed planar electrical imaging sensor. This imaging sensor acquires conductivity images of the liquid at the rear channel wall with a speed of 2,500 frames/s. By adding NaCl tracer to the denser fluid we were able to visualize the mixing process in high spatial and temporal detail. Furthermore, an image processing algorithm based on the optical flow concept was implemented and tested which allows the measurement of flow pattern velocities. Selected experiments at different Richardson numbers were run with two components of different density (pure water and glucose-water mixture) simulating borated and deborated water in a light water reactor scenario. These experiments were compared to CFD calculations using standard turbulence models. Good agreement between experimental data and CFD simulations was found.     

RELAP5/MOD3.3 analysis of coolant depletion tests after safety injection failure during a large-break loss-of-coolant accident

This study is concerned with development of a coupled calculation methodology with which to continually and consistently analyze progression of an accident from the design-basis phase via core uncovery to core melting and relocation. Experiments were performed to investigate the core coolant inventory depletion after safety injection failure during a large-break loss-of-coolant accident in a cold leg utilizing the Seoul National University Facility (SNUF). The SNUF is an integral test loop scaled down to 1/6.4 in length and 1/178 in area from the Advanced Power Reactor 1400 MWe (APR1400). The SNUF tests are simulated with the RELAP5/MOD3.3 code. The test results revealed that the core coolant inventory decreased five times faster during the sweepout in the downcomer than after termination of the sweepout. The sweepout was observed to take place on top of spillover from the downcomer region to expedite the depletion of the core coolant inventory. The calculation results of RELAP5/MOD3.3 deviated from the experimental data in terms of entrainment from the surface of core coolant, condensation and sweepout in the downcomer. Thereby, the core coolant level was computed to decrease faster than the measured from the experiment due to the overestimated spillover by the evaporation of the entrained droplets by the uncovered heaters. Notwithstanding the occasional disparities, the code prediction is in reasonable agreement with the overall behavior of the tests.     

Fluid mixing and flow distribution in the reactor circuit, measurement data base

Experimental investigations and computational fluid dynamics (CFD) calculations on coolant mixing in pressurised water reactors (PWR) have been performed within the EC project FLOMIX-R. The project aims 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. Measurement data from a set of mixing experiments have been gained by using advanced measurement techniques with enhanced resolution in time and space. Slug mixing tests simulating the start-up of the first main circulation pump are performed with two 1:5 scaled facilities: the Rossendorf Coolant Mixing model ROCOM and the Vattenfall test facility. Additional data on slug mixing in a VVER-1000 type reactor have been gained at a 1:5 scaled metal mock-up at EDO Gidropress. Experimental results on buoyancy driven mixing of fluids with density differences have been obtained at ROCOM and the Fortum PTS test facility.Concerning mixing phenomena of interest for operational issues and thermal fatigue, flow distribution data available from commissioning tests at PWRs and VVER are used together with the data from the ROCOM facility as a basis for the flow distribution studies.In the paper, the experiments performed are described, results of the mixing experiments are shown and discussed. Efforts on computational fluid dynamics codes validation on selected mixing tests applying Best Practice Guidelines in code validation will be reported about in a separate paper.     

Investigation of fluid mixing with direct vessel injection

The object of this work is to investigate fluid mixing phenomena as they related to pressurized thermal shock (PTS) in a pressurized water reactor vessel downcomer during transient cooldown with direct vessel injection (DVI), using test models. The test model designs were based on ABB Combustion Engineering (CE) System 80+ reactor geometry. A cold-leg, small-break loss-of-coolant accident (LOCA) and a main steam line break were selected as the potential PTS events for the ABB-CE System 80+. This work consists of two parts. The first part provides the visualization tests of the fluid mixing between DVI fluids and existing coolant in the downcomer region, and the second part presents the results of thermal mixing tests with DVI in the other test model. Flow visualization tests with DVI have clarified the physical interaction between DVI fluid and primary coolant during transient cooldown. A significant temperature drop was observed in the downcomer during the tests of a small-break LOCA. The measured transient temperature profiles compare well with the predictions from the REMIX code for a small-break LOCA, and with the calculations from the COMMIX-1B code for a stream line break event.     

HEXTRAN–SMABRE calculation of the VVER-1000 coolant transient-1 benchmark

The VVER-1000 coolant transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three-dimensional neutron kinetic core models. It concerns switching on a main coolant pump when the other three main coolant pumps are in operation. The problem is based on an experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, an extreme scenario concerning a control rod ejection after switching on a main coolant pump was calculated. At VTT the three-dimensional advanced nodal code HEXTRAN is used for the core dynamics, and the system code SMABRE as a thermal hydraulic model for the primary and secondary loop. The parallelly coupled HEXTRAN–SMABRE code has been in production use since early 1990s, and it has been extensively used for analyses of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used at VTT. The whole core calculation is performed with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation were specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Parametric studies have been performed for selected parameters.     

VAPEX-Code Analysis of a Melt–Coolant Interaction Experiment on the FARO Setup

The purpose of this work was to make a computational analysis using the VAPEX thermohydraulic code, developed at the Experimental Scientific-Research Center at the All-Russia Scientific-Research Institute of Nuclear Power Plants, of the L-33 experiment performed on the FARO setup to study the interaction of core melt with water. The calculation of the stage of mixing of the melt stream with water and subsequent cooling satisfactorily predicted experimental parameters such as the temperature of the water and gas in the vessel. The calculation of the formation and propagation of the thermodetonation wave agrees well with the experimental data. Analysis of the experiment showed that the VAPEX code can predict adequately the most important characteristics of melt–coolant interaction during a serious accident at a nuclear power plant in accordance with the out-of-vessel scenario of the development of the accident.