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

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

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.     

Investigation of Sweepout Mechanism and Critical Void Height in Annular Downcomer

The APR1400 (Advanced Power Reactor 1,400 MWe) has adopted the direct vessel injection (DVI) in lieu of the conventional cold leg injection for its emergency core cooling system (ECCS). In this reactor, sweepout from the water surface by gas (vapor or air) flow plays an important role in analyzing the mass and momentum transfer in the reactor downcomer of multidimensional geometry during a loss-of-coolant accident (LOCA) by decreasing the water level in the downcomer. The core water level will tend to decrease rapidly if a considerable amount of the entrained water stream and droplets bypasses through the break. The amount of entrained water is mostly determined by the interacting gas flow rate, the geometric condition, and the interfacial area between the gas and the water. The sweepout is observed to take place in three rather distinct steps: the beginning of undulation, the full wave and the wave peak (droplet separation). In view of these observations we investigated the relation between the gas flow rate and the amount of bypass as a function of time. The current experimental results shed light on the flow mechanism and the semi-empirical relations for the three-dimensional sweepout in a large-diameter annulus such as the reactor downcomer. A physico-numerical model is being developed to predict the multidimensional bypass flow rate resulting from the sweepout and entrainment in the downcomer.     

Direct vessel inclined injection system for reduction of emergency core coolant direct bypass in advanced reactors

Multidimensional thermal hydraulics in the APR1400 (Advanced Power Reactor 1400 MWe) downcomer during a large-break loss-of-coolant accident (LBLOCA) plays a pivotal role in determining the capability of the safety injection system. APR1400 adopts the direct vessel injection (DVI) method for more effective core penetration of the emergency core cooling (ECC) water than the cold leg injection (CLI) method in the OPR1000 (Optimized Power Reactor 1000 MWe). The DVI method turned out to be prone to occasionally lack in efficacious delivery of ECC to the reactor core during the reflood phase of a LBLOCA, however. This study intends to demonstrate a direct vessel inclined injection (DVII) method, one of various ideas with which to maximize the ECC core penetration and to minimize the direct bypass through the break during the reflood phase of a LBLOCA. The 1/7 scaled down THETA (Transient Hydrodynamics Engineering Test Apparatus) tests show that a vertical inclined nozzle angle of the DVII system increases the downward momentum of the injected ECC water by reducing the degree of impingement on the reactor downcomer, whereby lessening the extent of the direct bypass through the break. The proposed method may be combined with other innovative measures with which to ensure an enough thermal margin in the core during the course of a LBLOCA in APR1400.     

Commissioning of the ATLAS thermal-hydraulic integral test facility

KAERI recently constructed a new thermal-hydraulic integral test facility for advanced pressurized water reactors (PWRs) – ATLAS. The ATLAS facility has the following characteristics: (a) 1/2-height&length, 1/288-volume, and full pressure simulation of APR1400, (b) maintaining a geometrical similarity with APR1400 including 2(hot legs) × 4(cold legs) reactor coolant loops, direct vessel injection (DVI) of emergency core cooling water, integrated annular downcomer, etc., (c) incorporation of specific design characteristics of OPR1000 such as cold leg injection and low-pressure safety injection pumps, (d) maximum 10% of the scaled nominal core power. The ATLAS will mainly be used to simulate various accident and transient scenarios for evolutionary PWRs, OPR1000 and APR1400: the simulation capability of broad scenarios including the reflood phase of a large-break loss-of-coolant accident (LOCA), small-break LOCA scenarios including DVI line breaks, a steam generator tube rupture, a main steam line break, a feed line break, a mid-loop operation, etc. The ATLAS is now in operation after an extensive series of commissioning tests in 2006.     

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.     

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.     

An Experimental Study of Thermal-hydraulic Phenomena in the Downcomer with a Direct Vessel Injection System of APR1400 during the LBLOCA Reflood Phase

The direct vessel injection (DVI) mode is adopted as a safety injection system in the place of a conventional cold leg injection (CLI) mode in the Advanced Power Reactor 1400MW (APR1400). It is expected that “sweep-out” and “direct ECC (Emergency Core Cooling) water bypass” are two most important bypass mechanisms of ECC water injected through the DVI lines during the LBLOCA reflood phase. Using the test facility of plane-channel type scaled down to 1/7 ratio of the prototype reactor (APR1400), we carry out the following tests for the investigation of the two mechanisms: water film spreading test, sweep-out test, and direct ECC water bypass test. From the water film spreading test, it was found that the curvature effect is negligible and the present modified linear scaling law is more appropriate than the linear scaling law. In the sweep-out test, the continuous onset is used to analyze the water height in the downcomer and the amount of ECC water bypass by sweep-out is compared with the previous correlations. The direct ECC water bypass test is performed in order to understand the ECC water film behavior in the downcomer.     

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.     

Scaling methodology for a reduced-height reduced-pressure integral test facility to investigate direct vessel injection line break SBLOCA

A scaling methodology for a small-scale integral test facility was investigated in order to analyze thermal-hydraulic phenomena during a DVI (direct vessel injection) line SBLOCA (small break loss-of-coolant accident) in an APR1400 (advanced power reactor 1400 MWe) pressurized water reactor. The test facility SNUF (Seoul National University Facility) was utilized as a reduced-height and reduced-pressure integral test loop. To determine suitable test conditions for simulating the prototype in the SNUF experiment, the energy scaling methodology was propose to scale the coolant mass inventory and the thermal power for a reduced-pressure condition. The energy scaling methodology was validated with a system code (MARS) analysis for an ideally scaled-down SNUF model and that predicted a reasonable transient of pressure and coolant inventory when compared to the prototype model. For the actually constructed SNUF, the effect of scaling distortions in the test facility's thermal power and the loop geometry was analytically investigated. To overcome the limitation of the thermal power supply in the facility, the convective heat transfer between primary and secondary systems at the steam generator U-tubes was excluded and a modified power curve was applied for simulating the core decay heat. From the code analysis results for the actual SNUF model, the application of the modified power curve did not affect the major events occurring during the transient condition. The results revealed that the scaling distortion in the actual SNUF geometry also did not strongly disturb significant thermal-hydraulic phenomena such as the downcomer seal clearing. Thus, with an adoption of the energy scaling methodology, the thermal-hydraulic phenomena observed in the SNUF experiment can be properly utilized in a safety analysis for a DVI line break SBLOCA in the APR1400.     

CAP1400反应堆压力容器下降段内气-液逆向流试验研究

为研究真实工况下CAP1400反应堆压力容器下降段气-液逆向流现象,以CAP1400为原型,搭建压力容器下降段高度和直径比为1:1、60°切片的试验台架。试验工质为空气和水,试验研究了不同安注(DVI)供水量、不同气量的气-液两相流动和应急堆芯冷却剂(ECC)旁通现象。试验结果表明,DVI供水量相同时,随着供气量的增加,气-液逆向流现象明显,当质量流速达到4kg/s及以上时,安注水不能全部进入堆芯;Kutateladze经验关系式和UPTF经验关系式都与试验结果存在较大偏差,不适用于CAP1400压力容器下降段试验;基于试验数据,拟合了新的经验关系式,且通过比较有无DVI挡块的试验数据,验证了DVI挡块可以降低ECC旁通水量,增强安注能力。     

CFD analysis for thermal mixing in a subcooled water tank under a high steam mass flux discharge condition

A Computational Fluid Dynamics (CFD) analysis for a thermal mixing test was performed for 30 s to develop the methodology for a numerical analysis of the thermal mixing between steam and subcooled water and to apply it to Advanced Power Reactor 1400 MWe (APR1400). In the CFD analysis, the steam condensation phenomenon by a direct contact was simulated by the so-called condensation region model. Thermal mixing phenomenon in the subcooled water tank was treated as an incompressible flow, a free surface flow between the air and the water, and a turbulent flow, which are implemented in the CFX4.4. The comparison of the CFD results with the test data showed a good agreement as a whole, but a small local temperature difference was found at some locations. A sensitivity analysis was performed to find the reason of the temperature difference. The commercial CFD code of CFX4.4 together with the condensation region model can simulate the thermal mixing behavior reasonably well when a sufficient number of mesh distributions and a proper numerical method are selected.     

Air-water multidimensional impingement flow on vertical flat wall

The emergency core cooling (ECC) water is supplied from the direct vessel injection (DVI) system in the Advanced Power Reactor 1400 MWe (APR1400) during a postulated large-break loss-of-coolant accident (LBLOCA). The velocity of ECC water exceeds 10 m/s in the early high pressure phase of LBLOCA and then is decreased to 2-3 m/s in the late phase of reflood. During the injection the flow behavior exhibits a complex mode involving impingement, bypass, entrainment, sweepout and condensation in the reactor downcomer. There is currently no model to accurately simulate the local and complicated flow behavior in the APR1400 downcomer during a LBLOCA. This study is aimed at developing models for the water film flow and deformation, both of which are expected to sizably affect the other multidimensional flow characteristics in the downcomer. Experimental studies are conducted to benchmark the predictive model by furnishing the boundary conditions for the analysis resorting to the Accelerated Liquid Phase Hydrodynamics Apparatus (ALPHA) and the Kinetic Aerodynamic Physics Parallelepiped Apparatus (KAPPA). The Poisson equation and potential theory are applied to formulate the behavior of the water film and air flow. In both the experimental and numerical studies, the temperature-dependent thermodynamic properties and the reactor vessel curvature are neglected to render the problem at hand tractable. The model is found to reasonably describe the downward film flow behavior. The water film is developed in proportion to the initial injection velocity of the ECC water. The downward velocity of water film is increased with the heights of injection. Regarding the film deformation the calculated results tend to deviate from the experimental data as the injected air velocity is increased. The disagreement is attributed to limitations inherent in the two-dimensional treatment and point source approach.     

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。     

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

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

Scaling for the ECC bypass phenomena during the LBLOCA reflood phase

As one of the advanced design features of the APR1400 (Advanced Power Reactor), a direct vessel injection (DVI) system is adopted instead of the conventional cold leg injection (CLI) system. It is known that the DVI system greatly enhances the reliability of the emergency core cooling (ECC) system. However, there is still a dispute on its performance in terms of water delivery to the reactor core during the reflood period of a large-break loss-of-coolant accident (LOCA). Thus, experimental validation is underway. In this paper, a new scaling method, using the time and velocity reduced “modified linear scaling law”, is suggested for the design of a scaled-down experimental facility to investigate the direct ECC bypass phenomena in the PWR downcomer.     

Integral loop test and assessment of modified RELAP5/MOD3.3 for RCS coolant inventory during a LBLOCA

This study investigates experimentally and analytically the thermal hydraulic phenomena during the transition from design basis accident (DBA) to beyond-DBA, particularly, the depletion of core coolant inventory. To investigate the overall thermal hydraulic behavior after safety injection (SI) failure during a large-break loss-of-coolant accident (LBLOCA) in a cold leg, an integral loop test was performed at the Seoul National University Integral Test Facility (SNUF), which was scaled down to 1/6.4 in length and 1/178 in area from the advanced power reactor 1400 MWe (APR1400) according to the three-level scaling method. The plant condition at 200.0 s as the base case and those at 625.0 and 1950.0 s as test cases after the initiation of LBLOCA were applied as initial conditions in the experiments, respectively. The experimental results showed that the sweepout increased the coolant flow discharged to the break depending on the steam flow rate in intact cold legs and the initial downcomer coolant level and expedited the depletion of the core coolant inventory.In the meantime, since RELAP5/MOD3.3 uses the average properties of donor volume as those of its connected junction, this scheme causes the mass and the energy flux in a junction to be calculated incorrectly if significant phase separation occurs in the donor volume such as in the downcomer during the LBLOCA. The sweepout model was developed and implemented in RELAP5/MOD3.3 to improve its calculation of coolant inventory during the LBLOCA. To assess the applicability of the modified RELAP5/MOD3.3 to the actual system, the experiments in SNUF were simulated by both the original and the modified RELAP5/MOD3.3. The original one predicted the discharge flow rate at the break larger than that of the experiment. On the other hand, the modified one calculated the discharge flow rate more similar to that of the experiment than the original one did. As a result, the modified RELAP5/MOD3.3 reduced the coolant flow discharged to the break to delay the initiation time of heater heat-up after SI failure during LBLOCA in a cold leg. This improved RELAP5/MOD3.3 will support a more realistic thermal hydraulic analysis in an integrated analysis system.     

Sweepout model implementation in RELAP5/MOD3.3 to improve RCS coolant inventory calculation during a LBLOCA

According to the experiments of the Upper Plenum Test Facility (UPTF) and advanced power reactor 1400 MW_e (APR1400), the sweepout in the downcomer has been identified to play an important role in depleting the core coolant inventory during a Large-Break Loss-of-Coolant Accident (LBLOCA). In order to identify the sweepout mechanism and to estimate the amount of coolant discharged by sweepout, the separate-effect test was carried out in the plate type test apparatus, which was scaled down to 1/5 of the size of the APR1400 downcomer. In addition, the sweepout model was developed by correlating the experimental data on the critical void height and the discharge flow rate at the break to the values of analytically derived non-dimensional parameters. This model was implemented in RELAP5/MOD3.3 to improve its calculation of coolant inventory loss during a LBLOCA. To validate the modified RELAP5/MOD3.3 by implementing the sweepout model, the sweepout separate-effect test was simulated by both the original and the modified RELAP5/MOD3.3. The original one predicted the different discharge flow rates according to the node size of the donor volume, and these flow rates were larger than those of the experiment. On the other hand, the modified one calculated the discharge flow rate and the critical void height much more similar to those of the experiment than the original model did. In the future, the improved RELAP5/MOD3.3 adopted in an integrated analysis system will support a more realistic thermal hydraulic analysis.     

Two-phase natural circulation flow of air and water in a reactor cavity model under an external vessel cooling during a severe accident

As a part of a study on a two-phase natural circulation flow between the outer reactor vessel and the insulation material in the reactor cavity under an external reactor vessel cooling of the Advanced Power Reactor (APR) 1400, a Hydraulic Evaluation of Reactor cooling Mechanism by External Self-induced flow-HALF scale (HERMES-HALF) experiment has been performed by using the non-heating method of an air injection. This large-scale experiment uses a half-height and half-sector model of the APR1400. This experiment has been analyzed to verify and evaluate the experimental results by using the RELAP5/MOD3 computer code. The RELAP5/MOD3 results have shown that the water circulation mass flow rate is very similar to the experimental results of the HERMES-HALF, in general. Increases in the water inlet area and the water level in the reactor cavity lead to an increase in the water circulation mass flow rate. The effects of an air injection mass flow rate and the water outlet area on the water circulation mass flow rate are dependent on the water inlet area size. As the water outlet moves to a lower position, the water circulation mass flow rate increases slowly.     

Analysis of System Thermal Hydraulic Responses for Passive Safety Injection Experiment at ROSA-IV/Large Scale Test Facility

An experiment was conducted at the ROSA-IV/Large Scale Test Facility (LSTF) on the performance of a gravity-driven emergency core coolant (ECC) injection system attached to a pressurized water reactor (PWR). Such a gravity-driven injection system, though not used in the current-generation PWRs, is proposed for future reactor designs. The experiment was performed to identify key phenomena peculiar to the operation of a gravity injection system and to provide data base for code assessment against such phenomena. The simulated injection system consisted of a tank which was initially filled with cold water of the same pressure as the primary system. The tank was connected at its top and bottom, respectively, to the cold leg and the vessel downcomer. The injection into the downcomer was driven primarily by the static head difference between the cold water in the tank and the hot water in the pressure balance line (PBL) connecting the cold leg to the tank top. The injection flow was oscillatory after the flow through the PBL became two-phase flow. The experiment was post-test analyzed using a JAERI modified version of the RELAP5/MOD2 code. The code calculation simulated reasonably well the system responses observed in the experiment, and suggested that the oscillations in the injection flow was caused by oscillatory liquid holdup in the PBL connecting the cold leg to tank top.