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.     

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。     

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.     

Initial Thermal-Hydraulic Behaviors under Simultaneous ECC Water Injection into Cold Leg and Upper Plenum in a PWR-LOCA

Core thermo-hydrodynamic characteristics under the combined injection mode before and just after the beginning of bottom reflood of a PWR-LOCA were experimentally studied by performing three tests in Slab Core Test Facility simulating a full radius slab section of a PWR. Emergency core cooling water was simultaneously injected into the upper plenum and the intact cold leg. The subcooling and the radial distribution of the upper plenum injection water were the test parameters.

The core was cooled by falling water before the beginning of bottom reflood. However, the core was finally quenched by bottom reflood. Before the beginning of bottom reflood, the transients of water level in the lower plenum were different among three cases, that is, the water level was rapidly or gradually increased in the first and second cases, respectively, or remained below the bottom of core barrel in the third case. The bottom reflood was much delayed in the last two cases. Even under the conditions with large upper plenum injection rate of subcooled water and with steam escape through the lower plenum, continuous fall back was not observed but the subcooled water was intermittently supported by the upward steam flow generated in the core.     

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.     

Numerical investigation of radial mixing capabilities in strongly buoyancy-influenced vertical, turbulent channel flows

The flow behavior in the HDR downcomer during setting of the initial conditions for blowdown tests is investigated with the numerical simulation program for turbulent channel flows, TURBIT-3. This computer code is based on the complete 3-dimensional non-stationary basic equations for mass, momentum and heat. The subgrid scale models used for the turbulence structures not directly resolved by the grid are extended to take into account the buoyancy in the case of turbulent channel flow. The extended computer code is used to investigate how fast differences in temperature can be reduced, which are caused by inadequate mixing in the lower plenum during upward flow in the downcomer under conditions of mixed convection. It appears that, contrary to the computations neglecting the influences of buoyancy, the temperature differences are rapidly reduced already in the entrance zone of the downcomer. In this zone, local recirculation takes place in the cold region, which is quickly suppressed with increasing distance from the entrance by the intensification of the turbulence effects. A hot chimney extending through the whole downcomer cannot develop. Already at half level, the influence of buoyancy can be considered to be negligible in the downcomer which is assumed adiabatic. Under these conditions it should be possible in principle to set the enthalpy stratification by the planned layout of the experiment in the HDR-pressure vessel.     

大破口失水事故时冷热段同时安注反应堆堆芯会更安全

大破口失水事故时，安注系统由冷段注入的大量冷却剂从压力壳和吊兰之间的环形通道经破口流入安全壳，只有少量的冷却剂流入堆芯。如果把安注系统同时安装在冷段和热段同时进行安注，从热段注入的冷却剂带走了上腔室和堆芯内的较多热量而降低了上腔室内的压力，使冷段注入的冷却剂较容易流入堆芯。同时，从热段注入的部分冷却剂在上腔室内撞击在导向管上后，沿着导向管流入堆芯，堆芯得到的冷却剂比单一冷段安注时得到的冷却剂要多，堆芯会更安全     

UPTF experiment refined PWR LOCA thermal-hydraulic scenarios: conclusions from a full-scale experimental program

A total of 34 tests were performed at upper plenum test facility (UPTF, a 1:1 scale test facility) to investigate the thermal-hydraulic phenomena in a pressurized water reactor (PWR) primary system during end-of-blowdown, refill and reflood phases of a loss-of-coolant accident (LOCA). Separate effect tests as well as integral tests were carried out. After the completion of the program a summary of the basic findings from the full-scale tests is given, focusing on thermal-hydraulic issues related to: two-phase flow phenomena at the ECC injection ports for cold or hot leg injection; the ECC delivery into core area via the downcomer or the tie plate; entrainment de-entrainment phenomena during reflood (i.e. the “steam binding” and driving water head reduction problems).     

Verification study on alternative ECCS concepts for a PWR

An experimental study for alternative ECCSs for a PWR was performed with the ROSA-II facility. It was found through the tests that the combined injection of hot water into the upper plenum and cold water into the lower plenum accompanied by a low pressure coolant injection system in the hot legs is quite effective for core cooling through the whole period of a LOCA in the case of a cold leg break. The test results were compared with analytical results of the RELAP4J code. The code is found capable of estimating discharge flow behavior fairly well and can predict the overall fluid behavior in the tested method of the improved ECCSs. However, the calculated core heat transfer disagrees with the test data when the counter-cuurent flow of the two phases on the core is dominant.     

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.     

II. Steady state thermal measurements on the secondary side of a PWR steam generator (BUGEY-4 nuclear power plant)

Steady state thermal measurements, water and shell temperature, are made in the downcomer of the steam generator of the PWR BUGEY-4. The measurements show that the flow is of a swirling type with the degree of swirl being a function of the load (power). The angular temperature distribution of water and shell exhibits a minimum that rotates from the hot leg to the cold leg with increasing load. The results also admit the conclusion that the feedwater and recirculation water are well mixed in the cylindrical section of the downcomer. Measurements and computations of fluid and shell temperatures are in good agreement for the steam generator in a state of natural circulation. For the state of forced circulation, the agreement is poor when it is assumed that the flow in the downcomer is turbulent and the steam generator outer shell is well insulated. The agreement is excellent for turbulent flow but with air leakage, or infiltration, between the steam generator outer shell and its insulation. If the hypothesis of air infiltration is ruled out, agreement between measurements and computations is much improved when the flow in the downcomer is taken as a turbulent core flow with an attendant laminar boundary layer. The existence of a laminar boundary layer would require the flat plate transition Reynolds number be enhanced by a factor of 100. This enhancement could result from the combined effects of swirl and density gradient in the downcomer.     

600MW压水堆安注箱设计研究

本文用美国核管会热工水力程序TRACE和图形化建模软件SNAP,建立了600 MW两环路压水堆一回路和二回路热工水力系统分析模型,并对安注箱的各设计方案进行大破口失水事故(LBLOCA)模拟计算,通过对比各设计方案在LBLOCA事故下计算出的峰值包壳温度,研究安注箱在大破口失水事故工况下的安注性能,最后给出了优化的设计方案,并提出了可行的设计改进建议。研究结果表明,上腔室和下降段同时注入的方式较冷段注入和下降段注入更有效,且恰当地选取初始安注箱压力,可有效降低峰值包壳温度,提高LOCA裕量。     

压水堆下腔室流场对堆芯流场和温度场影响数值分析

利用三维数值模拟,对不同环腔厚度和环腔内冷却剂速度条件下,下腔室内冷却剂流场进行了计算。在此基础上,对压水堆流量孔板冷却剂流量的分配情况进行了分析,并找出了通过流量孔板通孔小组的冷却剂流量与平均流量的最小偏差。分别计算了最小偏差条件下与平均流量条件下,堆芯内板状燃料元件周围冷却剂的流场和温度场。发现由环腔厚度或环腔内冷却剂速度不同而引起的下腔室流场分布不均匀,对堆芯内冷却剂流场和温度场影响较大。     

One-dimensional three-field model of condensation in horizontal countercurrent flow with supercritical liquid velocity

A one-dimensional three-field model was developed to predict the flow of liquid and vapor that results from countercurrent flow of water injected into the hot leg of a PWR and the oncoming steam flowing from the upper plenum. The model solves the conservation equations for mass, momentum, and energy in a continuous-vapor field, a continuous-liquid field, and a dispersed-liquid (entrained-droplet) field. Single-effect experiments performed in the upper plenum test facility (UPTF) of the former SIEMENS KWU (now AREVA) at Mannheim, Germany, were used to validate the countercurrent flow limitation (CCFL) model in case of emergency core cooling water injection into the hot legs. Subcooled water and saturated steam flowed countercurrent in a horizontal pipe with an inside diameter of 0.75 m. The flow of injected water was varied from 150 kg/s to 400 kg/s, and the flow of steam varied from 13 kg/s to 178 kg/s. The subcooling of the liquid ranged from 0 K to 104 K. The velocity of the water at the injection point was supercritical (greater than the celerity of a gravity wave) for all the experiments. The three-field model was successfully used to predict the experimental data, and the results from the model provide insight into the mechanisms that influence the flows of liquid and vapor during countercurrent flow in a hot leg. When the injected water was saturated and the flow of steam was small, all or most of the injected water flowed to the upper plenum. Because the velocity of the liquid remained supercritical, entrainment of droplets was suppressed. When the injected water was saturated and the flow of steam was large, the interfacial shear stress on the continuous liquid caused the velocity in the liquid to become subcritical, resulting in a hydraulic jump. Entrainment ensued, and the flow of liquid to the end of the hot leg was greatly reduced.The influence of condensation on the transition from supercritical to subcritical flow as observed in the experimental data is also predicted with the three-field model. When the injected water was subcooled, condensation on the flow of continuous liquid caused a reduction in the flow of vapor and, consequently, a reduction in the interfacial shear stress. Therefore, the flow of liquid remained supercritical to the end of the hot leg at the upper plenum. The entire flow of injected water flowed to the end of the hot leg at higher flows of steam when the injected water was subcooled than when it was saturated. When the flow of vapor was large enough to cause a hydraulic jump in the subcooled liquid, the rate of entrained droplets was greatly increased. The interfacial surface area of the droplets was several orders of magnitude greater than for the continuous-liquid field, and condensation rate on the droplet field was also several orders of magnitude greater. When the flow of vapor from the upper plenum was at its greatest, most of the flow in the continuous liquid was entrained before reaching the upper plenum. The large flow of subcooled droplets caused three-quarters of the steam to condense.     

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。     

Multidimensional two-phase flow regime distribution in a PWR downcomer during an LBLOCA refill phase

The multidimensional countercurrent two-phase flow regimes that occur in a pressurized-water reactor (PWR) vessel downcomer during the refill phase of a large-break loss-of-coolant accident are studied using a transparent 1/10 scale model of a PWR vessel. The various flow regimes and their distribution in the downcomer have been identified and mapped for a range of air-water flooding experiments. The two-phase flow patterns that are identified in the downcomer included various types of film flows, droplet flows, countercurrent churn flows and cocurrent flows depending on the flooding condition. Through observation of the two-phase flow dynamics it was deduced that the physical mechanisms associated with the flooding processes could be separated into a liquid entrainment process and a film flow reversal process. In addition to the above exercise, the effect of non-uniform injection of water into the downcomer via different combinations of cold leg was studied similarly by determining flooding curves and flow pattern maps. It was found that differences in the flooding characteristic were noticeable for various water inlet configurations when compared with the uniform injection case. The differences could be explained qualitatively in terms of the flooding mechanisms identified previously by examining the flow patterns in the downcomer for the non-uniform injection tests.     

Measurement of turbulent flow phenomena for the lower plenum of a prismatic gas-cooled reactor

Mean velocity field and turbulence data are presented that measure turbulent flow phenomena in an approximately 1:7 scale model of a region of the lower plenum of a typical prismatic gas-cooled reactor (GCR) similar to a General Atomics design (Gas-Turbine-Modular Helium Reactor). The data were obtained in the Matched-Index-of-Refraction (MIR) facility at Idaho National Laboratory (INL) and are offered as a benchmark for assessing computational fluid dynamics (CFD) software. This experiment has been selected as the first Standard Problem endorsed by the Generation IV International Forum. The primary objective of this paper is to document the experiment and present a sample of the data set that has been established for this standard problem.Present results concentrate on the region of the lower plenum near its far reflector wall (away from the outlet duct). The flow in the lower plenum consists of multiple jets injected into a confined crossflow—with obstructions. The model consists of a row of full circular posts along its centerline with half-posts on the two parallel walls to approximate flow scaled to that expected from the staggered parallel rows of posts in the reactor design. Posts, side walls and end walls are fabricated from clear, fused quartz to match the refractive index of the mineral oil working fluid so that optical techniques may be employed for the measurements. The benefit of the MIR technique is that it permits optical measurements to determine flow characteristics in complex passages and around objects to be obtained without locating intrusive transducers that will disturb the flow field and without distortion of the optical paths. An advantage of the INL system is its large size, leading to improved spatial and temporal resolution compared to similar facilities at smaller scales. A three-dimensional (3D) particle image velocimetry (PIV) system was used to collect the data. Inlet-jet Reynolds numbers (based on the hydraulic diameter of the jet and the time-mean average flow rate) are approximately 4300 and 12,400. Uncertainty analysis and a discussion of the standard problem are included.The measurements reveal complicated flow patterns that include several large recirculation zones, reverse flow near the simulated reflector wall, recirculation zones in the upper portion of the plenum and complex flow patterns around the support posts. Data include three-dimensional PIV images of flow planes, data displays along the coordinate planes (slices) and presentations that describe the component flows at specific regions in the model.     

Investigation of Break Location Effects on Thermal-Hydraulics during Intermediate Break Loss-of-Coolant Accident Experiments at ROSA-III

The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR/6) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. Break location effects on thermal-hydraulics during intermediate LOCAs were investigated by using four experiments at the ROSA-III, the 15 and 25% main recirculation pump suction line break (MRPS-B) experiments, the 21% single-ended jet pump drive line break (JPD-B) experiment and the 15% main steam line break (MSL-B) experiment. Water injection from the high pressure core spray (HPCS) was not used in any of the experiments. Failure of ECCS actuation by the high containment pressure was also assumed in the tests.

In the MRPS-B experiments, the discharge flow turned from low quality fluid to high quality fluid when the downcomer water level dropped to the main recirculation line outlet elevation, which suppressed coolant loss from the vessel and the core. In the JPD-B experiment, the jet pump drive nozzle was covered with low quality fluid and low quality fluid discharge continued even after the downcomer water level reached the jet pump suction elevation. Low quality fluid discharge ceased after the ADS actuation. It suggestes that the JPD-B LOCA has the possibility of causing larger and more severe core dryout and cladding temperature excursion than the MRPS-B LOCA. The MSL-B LOCA was characterized by mixture level swell in the downcomer and the core. The core mixture level swell resulted in the much later core dryout initiation than that in the MRPS-B LOCA, however, ECCS actuation was also delayed because of slow downcomer water level drop.     

Upper Plenum Dump during Reflood in PWR Loss-of-Coolant Accident

Upper plenum dump during reflood in a large break loss-of-coolant accident of PWR is studied with the emergency core coolant injection into the upper plenum in addition to the cold leg. Transient experiments were carried out by injecting water into the upper plenum and the simple analysis based on a one-dimensional model was done using the drift flux model in order to investigate the conditions under which water dump through the core occurs during reflood.

The most significant result is an upper plenum dump occurs when the pressure (hydrostatic head) in the upper plenum is greater than that in the lower plenum. Under those circumstances the flow regime isco-current down flow in which the upper plenum is rapidly emptied. On the other hand, when the upper plenum pressure (hydrostatic head) is less than the lower plenum pressure (hydrostatic head), the co-current down flow is not realized but a counter-current flow occurs. With subcooled water injection to the upper plenum, co-current down flow is realized even when the upper plenum hydrostatic head is less than the lower plenum hydrostatic head. The importance of this effect varies according to the magnitude of water subcooling.     

Coolant flow field in a real geometry of PWR downcomer and lower plenum

Nuclear vendors and utilities perform numerous simulations and analyses in order to ensure the safe operation of nuclear power plants (NPPs). In general, the simulations are carried out using vendor-specific design codes and best-estimate system analysis codes, most of which were developed based on one-dimensional lumped parameter models. During the past decade, however, computers, parallel computation methods, and three-dimensional computational fluid dynamics (CFD) codes have been dramatically enhanced. The use of advanced commercial CFD codes is considered beneficial in the safety analysis and design of NPPs. The present work analyzes the flow distribution in the downcomer and lower plenum of Korean standard nuclear power plants (KSNPs) using STAR-CD. The lower plenum geometry of a PWR is very complicated since there are so many reactor internals, which hinders in CFD analysis for real reactor geometry up to now. The present work takes advantage of 3D CAD model so that real geometry of a PWR is used. The results give a clear figure about flow fields in the downcomer and lower plenum of a PWR, which is one of major safety concerns.