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.     

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.     

Ex-vessel transport and mixing of a deborated slug in a PWR primary geometry

The transport and mixing of a slug of deborated water in a lowered loop PWR is modeled by partitioning the volumes of the primary system according to chemical rector theory. Piping is modeled as plug flow volumes while the steam generator outlet plenum and the reactor coolant pumps are modeled as backmixed volumes. This simple approach provides a good representation of the transport and mixing phenomena outside the reactor vessel. The proposed methodology can be used to generate initial and boundary conditions for separate effects tests and CFD computations for the reactor vessel complex geometry. The decoupling of the ex-vessel primary system greatly enhances the resolution of boron dilution transient issue.     

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.     

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.     

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.     

压力容器环形下降段内含硼溶液流动混合特性的数值模拟分析

发生堆芯应急冷却安注时,外部注入的含硼冷却剂与稀释水团将在环形下降段内发生混合,含硼冷却剂与稀释水团混合不均匀可能导致堆芯重返临界。本文基于Fluent 18.0对环形下降段内的流动混合特性进行分析。横截面的速度分布显示,入口截面的水平方向速度随周向位置的增加而显著衰减,而环形下降段下部区域横截面的速度分布趋于平缓;三维流线图显示,流体进入压力容器后在环腔内壁发生剧烈碰撞,随后绕环形下降段呈放射状流动。通过自定义硼酸溶液,并模拟其与稀释水团之间混合,数值结果与相关的实验研究结果较为一致;三维浓度分布显示,雷诺数较低时入口硼酸溶液将停滞在环形下降段上部空间,增加入口雷诺数有利于搅混均匀。     

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.     

Dynamic Response of Hot/Cold Liquid Interfaces to Pump Speed Perturbations in a Thermal-Hydraulic Loop Simulating a PIUS-type Reactor

The operation of a PIUS-type reactor requires controlling the reactor pump speed to keep stationary the hot/cold liquid interfaces between the reactor coolant and cold borated water. The dynamic response of the interface location to pump speed perturbations is analyzed for an experimental loop simulating a PIUS-type reactor. The transfer function between the pump speed and the interface location is obtained by perturbing and Laplace-transforming the one-dimensional fluid momentum equations. The analytical results agree well with experimental data taken from the same facility. It is shown that the magnitude of the phase lag in the response of the interface location, which needs to be considered in designing a pump speed controller, primarily depends on the fluid inertia in the loop, the density lock flow area, and the density difference between the simulated reactor coolant and borated water.     

反应堆压力容器下腔室交混特性的数值模拟方法研究

反应堆冷却剂系统蒸汽管道发生破口事故后,硼溶液在反应堆压力容器下腔室的对流交混特性对于反应堆安全分析及事故后缓解与抑制策略制定均有重要作用。本文基于实验结果分析了反应堆压力容器下腔室的交混特性及浓度扩散过程,采用数值模拟方法结合实验数据比较了几种主要模型计算结果的准确性与可靠性。分析结果表明,压力容器下腔室的交混特性呈现出外围扩散特征,温度梯度法与组分输运模型具备描述浓度梯度扩散过程的能力,但在细节分布上仍存在进一步改善与优化的空间。     

Air-ingress analysis: Part 2—Computational fluid dynamic models

Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy (DOE), is performing research and development that focuses on key phenomena important during potential scenarios that may occur in very high-temperature reactors (VHTRs). Phenomena identification and ranking studies to date have ranked an air-ingress event, following on the heels of a VHTR depressurization, as important with regard to core safety. Consequently, the development of advanced air-ingress-related models and verification and validation data are a very high priority.Following a loss of coolant and system depressurization incident, air will enter the core of the high-temperature gas-cooled reactor through the break, possibly causing oxidation of the core and reflector graphite structure. Simple core and plant models indicate that, under certain circumstances, the oxidation may proceed at an elevated rate with additional heat generated from the oxidation reaction itself. Under postulated conditions of fluid flow and temperature, excessive degradation of lower plenum graphite because of oxidation might lead to a reactor safety issue. Computational fluid dynamics models developed in this study will improve our understanding of this phenomenon and is used to mitigate air ingress.This paper presents three-dimensional (3D) computational fluid dynamic (CFD) results for the quantitative assessment of the air-ingress phenomena. The 3D CFD simulation results show that the air-ingress accident is not controlled by molecular diffusion but density gradient driven stratified flow when the double-ended-guillotine break is assumed in a horizontal pipe configuration. It concludes that the previous air-ingress scenarios based on the molecular diffusion might not be correct and should be extensively modified to include real phenomena. This paper also presents a preliminary two-dimensional (2D) CFD simulation for validating an air-ingress mitigation concept using helium injection at the lower plenum. This simulation shows that the helium replaces air by buoyancy force and effectively mitigates air-ingress into the core.     

Simulating the mixing of helium and air by mixing salt water and fresh water in a scaled enclosure

We are attempting to establish scaling laws to simulate the mixing of helium (a simulant for hydrogen) with air in a large-scale enclosure by mixing salt water and fresh water, in a small-scale enclosure. This will allow us to assess the mixing of gases in a nuclear reactor containment using relatively small-scale liquid-mixing experiments. The scope of our current work does not cover the integrated effects of different mechanisms of gas mixing expected to prevail during postulated LOCA (loss-of-coolant accident) scenarios. The work is limited to mixing caused by jet inertia and buoyancy forces. Within this scope, we have identified the dominant scaling laws, and tested them by conducting gas-mixing experiments in a large-scale enclosure and liquid-mixing experiments in a small-scale enclosure. The experimental results demonstrate the validity of the scaling laws.     

Validation of the CFD code fluent by post-test calculation of a density-driven ROCOM experiment

During the last years, boron dilution events with the potential of reactivity transients were an important issue of German PWR safety analyses. A coolant with a low-boron concentration could be collected in localized areas of the reactor coolant system, e.g., by separation of a borated reactor coolant into highly concentrated and diluted fractions (inherent dilution) which can occur during reflux-condenser heat transfer after a small break loss of coolant accident with a limited availability of the emergency core cooling systems.During the course of follower core assessments, TÜV NORD SysTec appraises safety analyses of boron dilution events presented by the utilities. These analyses are based on the simulation of boron dilution and transport processes in conjunction with a number of dedicated experiments. The analyses demonstrate that boron dilution events cannot lead to recriticality of the core. Hence, the boron concentration at the core inlet has to be determined.TÜV NORD SysTec applies the CFD code FLUENT for the investigation of boron dilution events in pressurized water reactors. To affirm the FLUENT abilities for the simulation of boron dilution events, a validation against the ROCOM experiment T6655_21 with a density-driven coolant mixing was performed. This validation proves that FLUENT is able to appropriately simulate the effects of boron transport and dilution such as streaks of coolant with lower density in the downcomer. Deficits were identified in the simulation of fluid layering in the cold leg, which fortunately have a rather small influence on the predicted core inlet concentration. Therefore, the boron concentration in the reactor core can be determined with sufficient accuracy to solve the safety issue, regardless of the core becoming critical or not.     

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.     

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.     

Experimental study of the coupled thermo-hydraulic–neutronic stability of a natural circulation HPLWR

The HPLWR (high performance light water reactor) is the European concept design for a SCWR (supercritical water reactor). This unique reactor design consists of a three pass core with intermediate mixing plena. As the supercritical water passes through the core, it experiences a significant density reduction. This large change in density could be used as the driving force for natural circulation of the coolant, adding an inherent safety feature to this concept design. The idea of natural circulation has been explored in the past for boiling water reactors (BWR). From those studies, it is known that the different feedback mechanisms can trigger flow instabilities. These can be purely thermo-hydraulic (driven by the friction – mass flow rate or gravity – mass flow rate feedback of the system), or they can be coupled thermo-hydraulic–neutronic (driven by the coupling between friction, mass flow rate and power production). The goal of this study is to explore the stability of a natural circulation HPLWR considering the thermo-hydraulic–neutronic feedback. This was done through a unique experimental facility, DeLight, which is a scaled model of the HPLWR using Freon R23 as a scaling fluid. An artificial neutronic feedback was incorporated into the system based on the average measured density. To model the heat transfer dynamics in the rods, a simple first order model was used with a fixed time constant of 6 s. The results include the measurements of the varying decay ratio (DR) and frequency over a wide range of operating conditions. A clear instability zone was found within the stability plane, which seems to be similar to that of a BWR. Experimental data on the stability of a supercritical loop is rare in open literature, and these data could serve as an important benchmark tool for existing codes and models.     

Experimental CFD grade data for stratified two-phase flows

Stratified two-phase flows were investigated at two test facilities with horizontal test-sections. For both, rectangular channel cross-sections were chosen to provide optimal observation possibilities for the application of optical measurement techniques. In order to show the local flow structure, high-speed video observation was applied, which delivers the high-resolution in space and time needed for CFD code validation.The first investigations were performed in the Horizontal Air/Water Channel (HAWAC), which is made of acrylic glass and allows the investigation of air/water co-current flows at atmospheric pressure and room temperature. At the channel inlet, a special device was designed for well-defined and adjustable inlet boundary conditions. For the quantitative analysis of the optical measurements performed at the HAWAC, an algorithm was developed to recognise the stratified interface in the camera frames. This allows to make statistical treatments for comparison with CFD calculation results. As an example, the unstable wave growth leading to slug flow is shown from the test-section inlet. Moreover, the hydraulic jump as the quasi-stationary discontinuous transition between super- and subcritical flow was investigated in this closed channel. The structure of the hydraulic jump over time is revealed by the calculation of the probability density of the water level. A series of experiments show that the hydraulic jump profile and its position from the inlet vary substantially with the inlet boundary conditions due to the momentum exchange between the phases.The second channel is built in the pressure chamber of the TOPFLOW test facility, which is used to perform air/water and steam/water experiments at pressures of up to 5.0 MPa and temperatures of up to 264 °C, but under pressure equilibrium with the vessel inside. In the present experiment, the test-section represents a flat model of the hot leg of the German Konvoi pressurised water reactor scaled at 1:3. The investigations focus on the flow regimes observed in the region of the elbow and of the steam generator inlet chamber, which are equipped with glass side walls. An overview of the experimental methodology and of the acquired data is given. These cover experiments without water circulation, which can be seen as test cases for CFD development, as well as counter-current flow limitation experiments, representing transient validation cases of a typical nuclear reactor safety issue.     

Experiments on slug mixing under natural circulation conditions at the ROCOM test facility using high-resolution measurement techniques and numerical modeling

ROCOM is a four-loop test facility used for the investigation of coolant mixing in the primary circuit of pressurized water reactors. Recently, a new sensor was developed for an improved visualisation and quantification of the coolant mixing in the downcomer. This new sensor array spans a dense measuring grid and covers nearly the whole downcomer. In the presented work, special emphasis was given to the comparison of the data of this sensor with the results of calculations using the Computational Fluid Dynamics (CFD) code ANSYS CFX. A coolant mixing experiment during natural circulation conditions has been conducted. The underlying scenario of this experiment is based on a boron dilution scenario following a SBLOCA event. The corresponding CFD code solution has been obtained using the Best Practice Guidelines. All main effects observed in the measurement are described by the calculation. The detailed comparison reveals that the calculation underestimates the coolant mixing inside the reactor pressure vessel.The measurement data, boundary conditions of the experiment and facility geometry can be made available to other CFD code users for benchmarking.     

Development of an Advanced Startup Procedure for a PIUS-type Reactor

An advanced startup procedure for the PIUS-type reactor has been developed. The procedure is related to the way to isolate the primary loops from the borated reactor pool by establishing stable hot/cold water interfaces in the so-called density lock sections. The procedure starts with accumulating preheated water in the high points of the steam-generator-side legs. Then, by restarting the reactor coolant pumps, the primary loops can be isolated from the pool as the primary loops reaches a uniformly higher temperature than the pool water. The additional components required for this procedure are only a low-pressure grade heater and a pump of small capacities. Since the isolation is achieved with the density locks left open, the core shutdown and cooling capabilities by means of the natural circulation of borated water are maintained in case of any abnormal events during startup. The feasibility and the predictability of this procedure were investigated by running an experiment in a scaled single-loop facility and conducting an analysis using a one-dimensional model. Both in the experiment and in the analysis. the primary loop was successfully isolated from the pool.