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.     

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.     

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.     

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.     

反应堆压力容器下降段水-蒸汽CCFL实验与模型研究

为探究反应堆压力容器下降段在喷放末期冷段安注过程中的水-蒸汽逆流特性,建立下降段逆向流动限制(CCFL)模型,开展了基于压力容器模化本体的下降段CCFL实验研究以及建模分析。通过实验研究获得了不同入口安注水流量、安注水过冷度、堆芯蒸汽流量等条件下的下降段环腔内的安注特性数据,并基于实验数据进行了CCFL建模分析。结果表明,开始发生CCFL的蒸汽无量纲流速与入口安注水无量纲流速呈现正相关,基于无量纲流速建立的模型斜率与入口安注水无量纲流速呈现高度指数关联。本文建立了适用于从不发生CCFL至不完全CCFL,再到完全CCFL的下降段水-蒸汽气液逆流全过程预测模型。     

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.     

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.     

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.     

Three-dimensional analysis of flow characteristics on the reactor vessel downcomer during the late reflood phase of a postulated LBLOCA

A three-dimensional CFD analysis has been performed on the flow characteristics in the reactor vessel downcomer during the late reflood phase of a postulated large-break loss-of-coolant accident (LBLOCA), in order to validate the modified linear scaling methodology that was applied in the MIDAS test facility of Korea Atomic Energy Research Institute. The vertical and circumferential velocity similarities are numerically tested for the 1/1 and 1/5 linear scale models for the APR1400 reactor vessel downcomer. The effects of scale on flow patterns, pressure and velocity distributions, and the impinging jet behavior are analyzed with the FLUENT code. In addition, a simplified half cylinder model with a single emergency core cooling (ECC) nozzle is numerically tested to investigate the scale effect on the spreading width and break-up of ECC water film. The qualitative and quantitative results indicate that the 1/5 modified linear scale model of the reactor vessel downcomer would reasonably preserve the hydrodynamic similarity with APR1400.     

小型压水堆压力容器内部三维流场计算

In the reactor safety analysis process, it is important to obtain an accurate flow field inside the pressure vessel. Taking the small pressurized water reactor as the research object, the computational fluid dynamics (CFD) method was used to calculate and analyze the internal flow field of the reactor pressure vessel, and the fuel assembly flow distribution and the lower head mixing characteristics were obtained. The results show that the maximum flow distribution coefficient of the fuel assembly is 1.032, the minimum value is 0.934, and the overall flow distribution is characterized by “large in the middle and small in the edge” under the high-speed symmetrical inlet condition of the two pumps. The flow vortex of the lower head is enhanced, and the uneven distribution of the flow distribution of the fuel assembly is increased, under the high-speed asymmetric inlet condition of the pump. The minimum mixing factor of the coolant flow at the core inlet was calculated to be 0.022 due to the insufficient mixing characteristics of the lower head.     

小型压水堆压力容器内部三维流场计算

反应堆安全分析过程中,获得反应堆压力容器内部准确的流场至关重要。以小型压水堆为研究对象,运用计算流体力学(CFD)方法对反应堆压力容器内部流场进行计算分析,获得燃料组件流量分配和下封头混合特性。结果表明:两泵高速对称入口条件下,燃料组件流量分配系数最大值为1.032,最小值为0.934,且流量整体分布呈现"中间大、边缘小"的特点;一泵高速非对称入口条件下,下封头流动漩涡增强,燃料组件流量分配的不均性增大;下封头混合特性计算得到堆芯入口冷却剂流量混合因子最小值为0.022,下封头冷却剂混合能力不足。     

Effect of the yaw injection angle on the ECC bypass in comparison with the horizontal DVI

The comparison tests for the direct emergency core cooling (ECC) bypass fraction were experimentally performed with a typical direct vessel injection (DVI) nozzle and an ECC column nozzle having a yaw injection angle to the gravity axis. The ECC yaw injection nozzle is newly introduced to make an ECC water column in the downcomer region. The yaw injection angle of the ECC water relative to the gravity axis is varied from 0 to (±)90° stepped by 45°. The tests are performed in the air–water separate effect test facility (direct injection visualization and analysis (DIVA)), which is a 1/7.07 linearly scaled-down model of the APR1400 nuclear reactor. The test results show that (1) if the ECC water column is injected into the wake region which is induced by the hot leg blunt body in the downcomer annulus, the ECC bypass fraction is greatly reduced compared with the typical horizontal ECC injection which makes ECC film on the downcomer wall. At the same time, the ECC penetration toward the lower downcomer region becomes larger than those of a typical horizontal type of direct vessel injection on the downcomer wall vertically. (2) If the ECC water column is injected near the broken cold leg, the ECC water is directly bypassed. Thus, the ECC penetration fraction is greatly reduced compared with a typical film type of the horizontal ECC injection. (3) In order to minimize the ECC bypass fraction, the ECC water should be injected toward the wake region of the hot leg blunt bodies.     

压水堆下腔室流量分布数值分析

建立了压水堆下腔室流场的三维数值计算模型,计算了不同环腔厚度和环腔内冷却剂速度条件下,下腔室内冷却剂的流场,分析了环腔厚度和环腔内冷却剂速度对下腔室流向堆芯的流量分布的影响。入口速度不同或环腔厚度不同,在下腔内冷却剂流动形成漩涡的位置、大小和流动速度均会发生改变,导致通过流量孔板通孔的流量分布不同。入口速度较低时,流量孔板上所有通孔的流量分布比较均匀,在平均值附近波动,流量最高的通孔小组出现在边缘处;入口速度较高时,流量明显地呈现出中心高边缘低的特点。通孔小组的流量最大值随着环腔厚度增加由孔板的中心向边缘移动。     

Advanced DVI for ECC direct bypass mitigation

An ECC direct bypass fraction during a late reflood phase of a LBLOCA is strongly dependent on the characteristics of the cross flow and the geometrical configuration of a DVI in the downcomer of a pressurized light water reactor. The important design parameters of a DVI are the elevation, the azimuthal angle, and the separator to prevent a steam-water interaction. An ECC sub-channel to separate or to isolate an ECC water from a high-speed cross flow is one of the important design features to mitigate the ECC bypass phenomena. A dual core barrel cylinder as an ECC flow separator is located between a reactor vessel and a core barrel outer wall in the downcomer annulus. A new narrow gap between the core barrel and the additional dual core barrel plays the role of a downward ECC flow channel or an ECC flow separator in a high-speed cross flow field of the downcomer annulus. The flow zone around a broken cold leg in the downcomer annulus has the role of a high ECC direct bypass due to a strong suction force while the wake zone of a hot leg has the role of an ECC penetration. Thus, the relative azimuthal angle of the DVI nozzle from the broken cold leg is an important design parameter. A large azimuthal angle from a cold leg to a hot leg needs to avoid a high suction flow zone when an ECC water is being injected. The other enhancing mechanism of an ECC penetration is a grooved core barrel which has small rectangular-shaped grooves vertically arranged on the core barrel wall of the reactor vessel downcomer annulus. These grooves have the role for a generation of a vortex induced by a high-speed cross flow. Since the stagnant flow in a lateral direction and rotational vortex provides the pulling force of an ECC drop or film to flow down into the lower downcomer annulus by gravity, the ECC direct bypass fraction is reduced when compared to the current design of a smoothed wall. An open channel of grooves generates a stagnant vortex, while a closed channel of grooves creates an isolated ECC downward flow channel from a high-speed lateral flow. In this study, new design concepts for a dual core barrel cylinder, grooved core barrel, and a reallocation of the DVI azimuthal angle are proposed and tested by using an air-water 1/5 scaled air-water test facility. The ECC direct bypass reduction performances of the new design concepts have been compared with that of the standard type of a DVI injection. The azimuthal angle of the DVI nozzle from a broken cold leg varies from −15° to +52° toward a hot leg. The test results show that the azimuthal injection angle is an effective parameter to reduce the ECC direct bypass fraction. The elevation of the DVI nozzle is also an important parameter to reduce the ECC direct bypass fraction. The most effective design for reducing the ECC direct bypass fraction is a dual core barrel. The reduction fraction when compared to the standard DVI is about −30% for the dual core barrel while it is −15% for the grooved core barrel.     

Boiling water reactor fuel heatup during a loss of coolant while refueling

The design of nuclear power plants includes provisions for heat removal from the reactor core in the event there is a loss of reactor coolant while shut down. Boiloff from decay heat can lead to inventory reduction and fuel heatup if no coolant makeup is available. Certain decay heat removal system failures in boiling water reactors can drain the upper vessel and downcomer. This leaves the water inside the core shroud at the same level as the top of the jet pumps. This becomes the starting point from which further inventory reduction is possible through boiloff. This study investigated the core thermal response following such a scenario. A simple model of the core was used for analysis of this sequence. The goal of the analysis was to determine the time at which the water in the core would boil down and fuel heat up to a specified temperature (1256 K). It is this interval during which the operator can take action that will mitigate the transient.     

Experimental validation of the modified linear scaling methodology for scaling ECC bypass phenomena in DVI downcomer

In the direct vessel injection (DVI) system downcomer, the direct emergency core coolant (ECC) bypass is activated during the reflood phase of a large-break loss-of-coolant accident (LBLOCA) by the interaction between the downward-flowing liquid-film and the transverse gas flow. Direct ECC bypass is reportedly the major bypass mechanism of ECC, and various experiments have been performed to obtain detailed information about the ECC bypass in a DVI downcomer. These lead to a proposed new scaling methodology, named ‘modified linear scaling’, which is expected to preserve the phase distribution in the downcomer and the ECC bypass phenomena. In the present study, modified linear scaling was experimentally validated in air–water tests comprising Test 21-D of the upper plenum test facility (UPTF). The counterpart tests of UPTF Test 21-D were performed in 1/7.3 and 1/4.0 scale models of a UPTF downcomer, and the test results were compared with the experimental data of UPTF Test 21-D. Two important parameters of direct ECC bypass – the normalized liquid-spreading width on the downcomer wall and the direct ECC bypass fraction, which is the fraction of input water bypassed out the broken cold-leg – were considered in the validation. The comparison revealed that the scaling parameters of direct ECC bypass are well preserved in the prototype and reduced models, from which we conclude that the modified linear scaling methodology is appropriate for designing a reduced test facility and for a scaling analysis of direct ECC bypass in the reflood phase of an LBLOCA.     

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。     

Fluent analysis of a ROSA cold leg stratification test

One of the OECD ROSA project tests, investigating temperature stratification in the cold legs and the downcomer during ECCS water injection under single-phase natural circulation conditions was analysed with the FLUENT code. The guidance given in the “Best Practice Guidelines for the Use of CFD in Nuclear Reactor Safety Applications” of the OECD GAMA group was followed. Steady-state calculations were performed with the Standard k-?, the Realizable k-? and the Reynolds Stress Model, the last one being closest to the measured results. The calculations indicate the predominance of buoyancy effects in the cold leg caused by the density difference between cold and hot water, while in the test it seems, as if mixing between the cold plume and hot water would be the prevailing mechanism. It is shown that the temperature distribution in the downcomer is strongly influenced by correct modelling of the cold leg-downcomer connection. A model with an abrupt transition leads to the colder fluid flowing to the core barrel, while in the test it was flowing down along the vessel wall. Modelling the rounded transition of the ROSA facility shifts the cold stream towards the vessel wall.     

Experimental investigation of in-vessel mixing phenomena in a VVER-1000 scaled test facility during unsteady asymmetric transients

In-vessel turbulent mixing phenomena affect the time and space distribution of coolant properties (e.g., boron concentration and temperature) at the core inlet which impacts consequently the neutron kinetics response. For reactor safety evaluation purposes and to characterize these phenomena it is necessary to set and validate appropriate numerical modelling tools to improve the current conservative predictions. With such purpose, an experimental campaign was carried out by OKB Gidropress, in the framework of the European Commission Project “TACIS R2.02/02 - Development of safety analysis capabilities for VVER-1000 transients involving spatial variations of coolant properties (temperature or boron concentration) at core inlet”. The experiments were conducted on a scaled facility representing the primary system of a VVER-1000 including a detailed model of the Reactor Pressure Vessel with its internals. The simulated transients involved perturbations of coolant properties distribution providing a wide validation matrix. The main achievements of the set of experiments featuring transient asymmetric pump behaviour are presented in this paper. The potential of the obtained experimental database for the validation of thermal fluid dynamics numerical simulation tools is also discussed and the role of computational fluid dynamics in supporting the experimental data analysis is highlighted.     

AC600全压堆芯补水箱补水实验研究

全压堆芯冰箱（ＣＭＴ）是ＡＣ６００压水堆非能动高压安注系统的主要设备。全压堆芯补水箱补水实验主要研究中,小破口失水事故时ＣＭＴ的重力排放特性,为验证安全分析计算机程序试验数据,中国核动力研究院建造了ＣＭＴ补水实验装置,并在该装置上模拟反应堆主管道中,小破口失水事故动态工况,完成了ＣＭＴ补水实验,本文给出了小破口失水事故工况堆芯水箱补水试验结果与分析。,