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.     

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.     

基于激光诱导荧光法的空泡份额测量

激光诱导荧光技术作为可视化实验领域中的新技术,在核反应堆热工水力研究中得到了越来越广泛的运用。将激光诱导荧光技术应用于气液两相流空泡份额的测量中,介绍了激光诱导荧光法测量空泡份额的原理,描述了激光诱导荧光法测量空泡份额的具体实施方法,说明了图像数据的处理方法,并将该测量空泡份额方法用于气液两相流实验测量研究,在不同工况下进行了竖直通道内两相流空泡份额测量实验。实验结果表明,利用激光诱导荧光技术测得的空泡份额与理论预测结果符合较好。运用该方法能对流场内的空泡份额分布进行连续测量,且不会对流场造成干扰。     

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.     

Boron mixing in complex geometries: flow structure details

Quantitative measurements of scalar dilution transients within a scaled model of a PWR downcomer are presented. The injection transients correspond to the case of a single pump start-up injecting into an initially stagnant system. Detailed results of the flow path and velocity field are presented for a nominal case where a neutrally buoyant front of dilute water is injected into the downcomer. Additional tests are made for a finite-sized slug of dilute water with a volume approximately equal to that of the downcomer and the lower plenum. The results of the tests indicate that the flow is dominated near the injection leg by an impinging jet, which is bifurcated into two primary streams by the cylindrical geometry of the downcomer. The two streams exit the downcomer on the opposite side of the injection leg. The splitting of the streams creates a bell-shaped recirculation region in the lower downcomer beneath the injection leg, which is the last volume to be displaced by the injected fluid flowing up from the lower plenum. A third plume sometimes forms at the boundary between the streams and the recirculation region, introducing mixed fluid into parts of this relatively stagnant area at earlier times than is typically observed. The results are discussed in the context of computational fluid dynamics codes that are starting to be used to simulate the mixing phenomena encountered in such complex systems.     

棒束通道内温度场分布特性研究

可视化实验技术越来越多地被应用于核反应堆系统参数的测量,本文基于激光诱导荧光（LIF）技术的特点,介绍该技术的难点和解决方案,并对棒束通道定位格架下游稳态流和脉动流下温度分布进行了研究。结果显示,通过对系统光学特性和染色剂特性研究,可提高LIF技术的应用范围和测量精度。同时采用后处理技术,可获得更准确的温度场分布。通过对棒束通道定位格架下游全场温度进行测量,获得了稳态流和脉动流两种工况下温度的分布。定位格架能显著增强下游的流动搅混,提高换热能力。流速的波动也会对温度分布产生显著影响。研究表明,LIF技术可实现对棒束通道内流体温度分布的全场测量,根据温度分布特性研究可实现对定位格架性能的评价。     

Temperature Field Distribution Characteristics in Rod Bundle Channel

Visualized experimental techniques are increasingly used in the measurement of nuclear reactor system parameters. Based on the characteristics of laser induced fluorescence (LIF) technique, the difficulties and solutions of the LIF technique were introduced in this paper. And the temperature distributions downstream of spacer grid in rod bundle channel under the steady flow and pulsating flow were analyzed. The results show that the application range and measurement accuracy of LIF technique can be improved by studying the optical properties and dye characteristics. At the same time, post-processing technology can be used to obtain more accurate temperature field distribution. The full-field temperature distributions downstream of spacer grid in rod bundle channel under steady flow and fluctuating flow conditions were obtained. Spacer grid can significantly enhance flow mixing and improve heat exchange capacity. Temperature distribution is also affected by fluctuations in velocity. In summary, the LIF technique can achieve the full-field measurement of the temperature distribution in the rod bundle channel. According to the temperature distribution characteristics, the performance of the spacer grid can be evaluated.     

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.     

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.     

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.     

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。     

承压热冲击下压水堆压力容器壁面换热特性的数值模拟研究

为了研究压水堆因安注冷水直接注入反应堆压力容器下降环腔而导致的承压热冲击（PTS）热工水力问题，基于1∶10比例模型，应用计算流体力学商用软件FLUENT5．4进行了紊流流动换热的数值模拟分析，同时进行了常压传热实验研究。针对下降环腔折算流速0．5m/s，安注流速10m/s的典型工况，研究了压力容器下降环腔的壁面换热特性。通过分析下降环腔内的流动及混合特性，从流动机理上解释了压力容器内壁上准重接触点附近换热强烈的现象，并指出壁面换热强弱与近壁流体紊流脉动动能密切相关，为热冲击分析提供参考。     

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.     

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.     

Thermal and stability considerations for a supercritical water-cooled fast reactor with downward-flow channels during power-raising phase of plant startup

This paper describes study on the procedure of raising the reactor thermal power and the reactor coolant flow rate during the power-raising phase of plant startup for the supercritical water-cooled fast reactor (SWFR), which is selected as one of the Generation IV reactor concepts. Since part of the seed fuel assemblies and all the blanket fuel assemblies of the SWFR are cooled by downward flow, the feedwater from the reactor vessel inlet nozzle to the mixing plenum located below the core is distributed among these fuel assemblies and the downcomer. The flow rate distribution as the function of both the reactor thermal power and the feedwater flow rate, which are the design parameters for the power-raising phase, is obtained by the thermal hydraulic calculations. Based on the flow rate distribution, thermal analyses and thermal-hydraulic stability analyses are carried out in order to obtain the available region of the reactor thermal power and the feedwater flow rate for the power-raising phase. The criteria for the “available” region are the maximum cladding surface temperature (MCST) and the decay ratio of thermal-hydraulic stability in three “hot” channels; two seed assemblies with upward/downward flow and a blanket assembly. The effects of various heat transfer correlations and axial power distributions are also studied.     

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.     

Simple mixing model for pressurized thermal shock applications

The phenomenon of fluid/thermal mixing in the cold leg and downcomer of a Pressurized Water Reactor (PWR) has been a critical issue related to the concern of pressurized thermal shock. The question of imperfect mixing arises when the possibility of cold emergency core cooling water contacting the vessel wall during an overcooling transient could produce thermal stresses large enough to initiate a flaw in a radiation embrittled vessel wall. The temperature of the fluid in contact with the vessel wall is crucial to a determination of vessel integrity since temperature affects both the stresses and the material toughness of the vessel material. A simple mixing model is described which was developed as part of the EPRI pressurized thermal shock program for evaluation of reactor vessel integrity.     

基于RELAP5的LOCA喷放阶段下降段内CCFL特性研究

反应堆失水事故（LOCA）后下降段通道内形成的两相逆流状态极有可能引发汽-液逆向流动限制（CCFL）,不利于应急冷却水顺利进入堆芯,极大影响了核反应堆系统的安全性能。本研究基于RELAP5程序采用Wallis溢流关系式对UPFT实验装置进行建模并计算LOCA喷放阶段的下降段注水行为;通过对比下腔室蓄水量、下降段内压力及破口处蒸汽流量瞬态变化以验证模型的有效性,并对下降段通道内汽相速度场、液相体积分数分布特性进行分析。结果表明,由于下降段通道结构的三维特征引起的流动不均匀性影响了汽-液CCFL特性,随着蒸汽流量增大,在破口环路与下降段连接区域的压力梯度与向上流速度梯度越大,较少节点的划分方法很难真实反映下降段通道局部区域内汽-液溢流关系;在靠近破口的环路内注入的冷却水更难到达下腔室,而在远离破口环路的冷却水容易进入到下腔室;过热的蒸汽在流动过程中被冷却水冷却发生凝结现象,导致出口蒸汽流量小于进口蒸汽流量,且随着进口蒸汽流量的增大,凝结效应则随之减小。本研究所建立的模型与方法能够适用于LOCA喷放阶段下降段通道内的汽-液CCFL预测。      

Accident progression within the secondary containment boundary of large LMFBRs

This study deals with the sodium spillage phenomenon as it relates to accident energetics and containment integrity. Sodium spillage has been identified as an important issue for large LMFBRs because of the large inventory of sodium present and the potential for energetic accidents. Energetic core-disruptive events leading to slug impact could open leak paths in the reactor cover and vent sodium into the secondary containment. Sodium fires in the containment building could lead to pressurization and thermal stressing of the surrounding structure and jeopardize containment integrity. The potential consequences of such a scenario have prompted the development of analytical tools to quantify the spillage process.One of the primary concerns in assessing the integrity of secondary containment is the amount and velocity of sodium which may be ejected from the primary vessel. A parametric study has been performed, the purpose of which was to study the sensitivity of sodium spillage to accident energetics. Treatment of the spillage process was accomplished with the ICECO code employing a quasi-Eulerian method. A 1000 MWe reactor, with prescribed leak paths, is modelled and analyzed during the slug impact phase. Leak paths are assumed to exist as annular penetrations in the reactor cover and as a gap at the vessel-head junction. The behavior of sodium spillage is investigated under conditions of different accident energetics, various opening sizes, and multiple leak paths, with both stationary and moving reactor covers. The relative influence of short and long term spillage is also addressed.During the transient period immediately following slug impact it was found that spillage from annular penetrations in the reactor cover is only weakly sensitive to changes in slug velocity. The same conclusion applies to spillage from a fixed gap at the vessel-head junction. Significant sensitivity of spillage to accident energetics was seen only in cases of spillage from the vessel-head junction when the reactor cover was movable. The influence of slug impact on the motion of the reactor cover leads to the conclusion that sodium spillage is most sensitive to accident energetics inasmuch as it affects the size of the leak path.