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To investigate thermal-hydraulic phenomena during an intermediate-break loss-of-coolant accident (IBLOCA) and evaluate the effect of a direct vessel injection (DVI) line break, an integral effect t...

To improve safety analysis technology for a nuclear reactor containment considering an interaction between a reactor coolant system (RCS) and containment, this study aims at an experimental investigation on the integrated simulation of the RCS and containment, with an integral effect test facility, ATLAS-CUBE. For a realistic simulation of a pressure and temperature (P/T) transient, the containment simulation vessel was designed to preserve a volumetric scale equivalently to the RCS volume scale of ATLAS. Three test cases for a steam line break (SLB) transient were conducted with variation of the initial condition of the passive heat sink or the steam flow direction. The test results indicated a stratified behavior of the steam-gas mixture in the containment following a high-temperature steam injection in prior to the spray injection. The test case with a reduced heat transfer on the passive heat sink showed a faster increase of the P/T inside the containment. The effect of the steam flow direction was also investigated with respect to a multi-dimensional distribution of the local heat transfer on the passive heat sink. The integral effect test data obtained in this study will contribute to validating the evaluation methodology for mass and energy (M/E) and P/T transient of the containment.

An experimental simulation for a loss of coolant accident with a 4-in cold leg top-slot break was performed using ATLAS, an integral effect test facility at the Korea Atomic Energy Research Institute (KAERI), South Korea. The aims of this test are to resolve a safety issue and to validate the thermal hydraulic safety analysis codes for the activity of the 4th ATLAS-Domestic Standard Problem. The related safety issue for this test is that a loop seal clearing and loop seal reformation of APR1400 can lead to a core temperature excursion because of the characteristics of the loop seal geometry. In the experimental results, the core heater temperature slowly increased after loop seal reformation. This increase is attributed to the accumulated steam at the upper head. The accumulated steam increased the pressure and saturated the temperature in the core. The core heater temperature then increased due to the increased saturated temperature. This temperature increase, therefore, was not a core temperature excursion. Sensitivity analysis results using Multidimensional Analysis of Reactor Safety KINS Standard (MARS-KS), which is a best-estimate thermal hydraulic system analysis code, are also presented in this study. In the 4th ATLAS-Domestic Standard Problem, 15 participants performed calculations and sensitivity analyses using various thermal hydraulic safety analysis codes (MARS-KS, SPACE, RELAP5, and TRACE). From this exercise, several parameters that affect the calculation results are brought out. The sensitivity analysis results using MARS-KS with some of these parameters are also presented. The MARS-KS calculation results are similar to the experimental data. The effects of critical flow model, break line modeling, and fine node modeling are also discussed.

Station blackout (SBO) is one of the most important design extension conditions (DECs) because a total loss of the heat sink leads to a core uncovery or damage without any proper operator action. During a long transient of an SBO, a steam generator tube rupture (SGTR) accident can occur when a steam generator tube is exposed to a superheated steam flow. In this study, a prolonged SBO with an SGTR occurrence was experimentally investigated by performing an integral effect test with the ATLAS facility. For the transient simulation, the tube rupture was simulated when the core water level became below the top of the active core. As an accident management measure, the auxiliary feedwater was supplied to the intact steam generator when the maximum heater surface temperature started to show an excursion behavior. The experimental results showed that a single tube rupture could not sufficiently reduce the primary system pressure before the injection of the auxiliary feedwater. A delayed supply of the auxiliary feedwater after an excursion of the heater surface temperature successfully cooled the primary system until the end of the transient, where inflow of the coolant from the pressurizer contributed to a recovery of the coolant inventory in the core. The natural circulation flow characteristics in the primary system showed an oscillating behavior depending on the heat removal rate of the steam generators. This integral effect test data can be used to evaluate the prediction capability of safety analysis codes and identify code deficiencies in predicting an SBO transient with an SGTR occurrence.

This study reports an experiment to investigate the wetted wall fraction (WWF), which represents the shape of the continuous interface of the air–water stratified and wavy flow regimes, in horizontal and inclined pipes with an inner diameter of 40 mm. Using the experimental data, a semi-empirical model to predict the WWF was developed based on the energy balance for the liquid phase in separated flows. The model used a relationship between the WWF and the center of gravity of the liquid phase with a concave interface that can describe continuous changes to the flow regime from a stratified to an annular. The coefficients of the model were empirically determined based on a wide range of experimental data from the literature obtained in the air and various liquids covering a void fraction of greater than 0.79, a pipe with inner diameter in the range 24–150 mm, density differences varying from 812 to 1052 kg/m3, a range of liquid phase viscosity of 0.87–5.66 mPa s, surface tension ranging from 27.9 to 72.7 mN/m, a wide range of inclination angles from −27° for a downward flow to +3° for an upward flow, and gas and liquid Reynolds numbers based on a superficial velocity of up to 219,000 and 7500, respectively. This WWF model was tested using an extensive experimental database of the WWF and void fraction in the stratified and wavy flow regimes, and yielded the best agreement compared with existing models in the literature.

KAERI has been operating an integral effect test facility, Advanced Thermal–Hydraulic Test Loop for Accident Simulation (ATLAS), for accident simulations of advanced pressurized water reactors. As ...

PAFS (Passive Auxiliary Feedwater System) is one of the advanced safety features in the design of the APR+ (Advanced Power Reactor) nuclear power plant, in order to cool down the reactor coolant system without any external supply of electricity or water during an accident. The driving force of the PAFS is condensation inside horizontal tubes of PCHX (Passive Condensation Heat Exchanger). To improve the prediction capability for the condensation heat transfer in the PCHX, advanced wall condensation models which mechanistically consider the wall condensation with the different flow regimes in a horizontal channel were implemented into a thermal hydraulic safety analysis code, SPACE (Safety and Performance Analysis Code for Nuclear Power Plant). The models were evaluated by comparing to the experimental data from the PASCAL (PAFS Condensing Heat Removal Assessment Loop) test facility, which is a separate effect test to validate the cooling performance of the PCHX. Calculation results showed that the condensation model package (Ahn et al., 2014) presented the enhanced the prediction capability of the SPACE code in predicting the steam pressure and temperature under quasi-steady state conditions. Since this model package is composed of mechanistic models by considering the film condensation in the steam phase and the convection in the condensate liquid separately, it contributed to enhance the prediction capability of the SPACE code and reduce the conservatism in the prediction of the cooling capability of the PAFS. The SPACE code with improvement of the wall condensation model can be applied in predicting the cooling and operation capability of the PAFS more realistically in the safety analysis of APR+ nuclear power plant.

An RCP test facility has been constructed at KAERI. During the commissioning startup operation, one of thermowells was cracked due to the high-speed coolant velocity. To complete the startup operation, a modified design of the thermowells was proposed, and all the original thermowells were replaced by the modified ones. From an evaluation of the original and modified designs of the thermowells according to the recent PTC code, the original design did not satisfy the new or old requirements, and the modified design simply satisfied the old requirement. The cracked surface of the original one was owing to the fatigue stresses in transverse oscillations. Two of three modified thermowells were found to be bent after about a 4.1 × 10 8 fatigue cycle operation. Their tapered-shanks came in contact with the nozzle walls and each showed a rim of deformation on the root. The modified ones suffered the fatigue stresses in in-line oscillations and became bent in in-line direction. A new design of thermowells with a straight shape and shorter unsupported length satisfying the current PTC code requirement was developed and installed in the RCP test facility. For the new design, additional modal analysis based on the finite element method was performed to check the dynamic behaviors using the ANSYS code and it was found that the PTC code predicted more conservative frequencies for the evaluation of dynamic behaviors in thermowell design.

The counterpart test of the OECD-ATLAS project named A5.1 was performed using a thermal–hydraulic integral effect test facility, ATLAS. The target scenario for the A5.1 test was the LSTF SB-CL-32 test. The SB-CL-32 test is a 1% horizontal small break loss of coolant accident (SBLOCA) at the cold leg with cold leg injection of the emergency core cooling system (ECCS) and accident management (AM) action. The test objective is to investigate the thermal hydraulic phenomena during a cold leg SBLOCA such as core heat-up, loop seal clearing (LSC), and the effect of AM actions. To achieve the test objectives, scaling analysis between LSTF and ATLAS was conducted first. The detailed test conditions such as initial and boundary conditions and sequence of events were determined based on the scaling analysis result. The experimental result showed that overall thermal hydraulic behavior which can be expected in this kind of SBLOCA scenario was observed. However, significant differences were found in the A5.1 test result compared with the LSTF SB-CL-32 test result in the view point of the loop seal clearing phenomenon and the maximum heater rod surface temperature behavior. These differences result from the different design between two test facilities which are refer to the different prototype nuclear power plant. Thus, several test condition parameters were analyzed as the sensitivity study utilizing the thermal hydraulic system analysis code to check their effect on the system response. From the sensitivity analysis result, it was confirmed that the A5.1 test was well modeled as a counterpart test of LSTF and realized the thermal hydraulic behavior during SBLOCA transient. This test result can be utilized to evaluate the prediction capability of thermal hydraulic system analysis codes and to validate the scale-up capability of integrated effect test (IET) test data to the nuclear power plant condition.

The thermal–hydraulic integral effect tests were performed to investigate the effect of asymmetric active and passive secondary cooling on the primary cool-down performance during a prolonged station blackout. The A1.1 test was performed to simulate a prolonged station blackout with asymmetric secondary cooling through the delayed supply of auxiliary feedwater only to steam generator number 2 of ATLAS. The A1.2 test was performed to simulate a prolonged station blackout with asymmetric secondary cooling through the supply of passive auxiliary feedwater only to steam generator number 2 of ATLAS. In both tests, an asymmetric cooling pattern was clearly observed during the period when the auxiliary feedwater was supplied. The primary system pressure and temperature continuously decreased during the heat removal by active and passive auxiliary feedwater supply, which indicates that the active and passive auxiliary feedwater supply can efficiently remove the core decay heat during a prolonged station blackout transient.

To provide a passive cooling system for the reactor containment, Passive Containment Cooling System (PCCS) was adopted in the design of i-POWER nuclear power plant. This study focused on validation tests for condensation heat transfer of the PCCS heat exchanger using the CLASSIC facility. The tests include investigation of the condensation heat transfer in prototypic single tube and bundle tubes. From the single tube experiments, condensation heat transfer model was proposed to reflect the PCCS heat exchanger tube geometry. Experimental results in the bundle tube show consistent trend compared to the proposed heat transfer model from the single tube test. The local condensation heat transfer coefficient of inside tubes was smaller than the average value due to a shadow effect by a larger mass fraction of non-condensable gas, so that design of the PCCS should take into account degradation of the condensation heat removal in the bundle geometry.

Main purposes of the LOCAL (LOop seal Clearance separate effect test of ATLAS) facility are to investigate the effect of cross-over leg (COL) diameter at the loop seal clearance (LSC), to evaluate the two-phase flow regime through high speed visualization, and to suggest the mechanistic models for the LSC prediction at Small Break-Loss of Coolant Accident (SB-LOCA) in Light Water Reactors (LWRs). In this paper, we introduce a key mechanism of the onset of the LSC occurring at U-shaped COL with air-water two-phase flow condition. The onset of LSC is coupled with the flow regime transition at vertical upward and horizontal section of COL. Slug flow with single Taylor bubble results in decrease of pressure difference at vertical upward section and causes rapid flow regime transition at horizontal section from the stratified to the slug. The critical velocity of gas leading to flow regime transition at horizontal section can explain the onset of LSC and high speed visualization can provide the clear understanding of LSC phenomena at the local point of view.

The integral effect test to simulate a steam line break (SLB) accident accompanied by a steam generator tube rupture (SGTR), named SLB-SGTR-02 test, was performed to investigate the thermal hydraulic phenomena during a multiple failure accident. In this study, a thermal-hydraulic integral effect facility, ATLAS (Advanced Thermal-hydraulic test Loop for Accident Simulation) was utilized to simulate thermal hydraulic phenomena which can occur in the nuclear power plant, especially for the pressurized water reactor, as realistically as possible. The SLB is simulated as a guillotine break on the main steam line and, in the SGTR, five steam generator tubes on the steam generator hot-side are modeled. Due to the initiation of SLB, a reactor was tripped by low steam generator pressure (LSGP) signal and the SGTR was initiated when the water inventory in the secondary system of the affected steam generator was dried-out. During the transient simulation for SLB-SGTR-02 test, major thermal-hydraulic parameters such as the system pressures, the collapsed water levels, the flows in the primary loops and the fluid temperatures, were measured and analyzed. Through this experimental result, insights about the accident management procedure can be provided in the case of the multiple failure accident, such as an SLB with a SGTR.

The US NRC’s post-Fukushima taskforce recommend extending the coping time up to at least 72 hours. In PWRs, the reactor coolant pump (RCP), and more particularly the shaft seal system located between the motor and the impeller, are directly impacted by these recommendations [AREVA, 2012]. Considering the importance of station black out (SBO) condition, seal performance of RCP shaft seal assembly shall be verified especially under an extended SBO condition. As a final stage of the localization program of the RCP, KAERI performed a seal LOCA test to verify seal leakage characteristics and to produce qualified experimental data for a plant safety analysis during the extended SBO condition. The seal verification test was performed using the localized seal assembly, fabricated by DOOSAN Heavy Industries & Construction Company of Korea, along with the commercial APR1400’s RCP for the Shin-Kori unit-5 nuclear plant to simulate an actual condition of the commercial nuclear power plant. Test condition was derived from a series of safety analysis for the APR1400 plant which reflected the FLEX procedure, and it consisted of three major sub-conditions such as high-pressure and high-temperature condition, saturated condition, and low-pressure and low-temperature condition. The major concerning point of the test is to measure a seal leakage flow rate during the test period. Total leakage amount and maximum leakage flow rate through the localized shaft seal assembly during the whole period of the test are in comparatively smaller ranges than the conventional input values used for a safety analysis of APR1400.

The integral effect test to simulate the safety injection pump (SIP) failure accompanied by a steam generator tube rupture (SGTR), named a SGTR-SIP-01 test, was performed to investigate the thermal hydraulic phenomena during a multiple failure accident. In this study, a thermal-hydraulic integral effect facility, ATLAS (Advanced Thermal-hydraulic test Loop for Accident Simulation) was utilized to simulate thermal hydraulic phenomenon which can be occurred in the nuclear power plant, as realistically as possible. In this SGTR-SIP-01 test, a rupture of five steam generator u-tubes on the steam generator hot-side was simulated. Due to the initiation of SGTR, a reactor was tripped by high steam generator level (HSGL) signal. During the transient simulation for SGTR-SIP-01 test, major thermal-hydraulic parameters such as the system pressures, the collapsed water levels, the flows in the primary loops, and the fluid temperatures, were measured and analyzed. Through this experimental result, insights about the accident management procedure can be provided in the case of the multiple failure accident, such as a SGTR accident with a total failure of SIPs. In addition to that, for improvement of the system code which are now on developing such SPACE or MARS-KS code, this test data can be utilized for validation and verification work.

In the high convective gas flow condition, irregular shaped water waves from which droplet entrainment occurs are generated under horizontally stratified two-phase flow condition. KAERI proposed a new mechanistic droplet entrainment model based on the momentum balance equation consisting of the shear stress, surface tension, and gravity forces. However, this model requires correlation or experimental data of several physical parameters related to the wave characteristics. In the present study, we tried to measure the physical parameters such as wave slope, wave hypotenuse length, wave velocity, wave frequency, and wavelength experimentally. For this, an experiment was conducted in the horizontal rectangular channel of which width, height, and length are, respectively, 40 mm, 50 mm, and 4.2 m. In the present test, the working fluids are chosen as air and water. The PIV technique was applied not only to obtain images for phase interface waves but also to measure the velocity field of the water flow. Additionally, we developed the parallel wire conductance probe for the confirmation of wave height from PIV image. Finally, we measured the physical parameters to be used in the validation of new droplet entrainment model.

KAERI (Korea Atomic Energy Research Institute) will launch an OECD/NEA project by utilizing a thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) from April, 2014. In the OECD-ATLAS project, design extension conditions (DECs) such as a station blackout (SBO) and a total loss of feed water (TLOFW) will be experimentally investigated to identify the major thermal-hydraulic phenomena of high risk multiple failure accidents considering magnified safety concerns after the Fukushima accident. A SBO is one of the most important DECs in that without any proper operator actions, a total loss of heat sink leads to core uncover, to core damage, and ultimately a core melt-down scenario under high pressure. Due to this safety importance, a SBO is considered to be a base test item of the OECD-ATLAS project. In this study, a phenomena identification ranking table (PIRT) has been developed for identifying the major parameters affecting the thermal-hydraulic phenomena in SBO transients. Development of a PIRT for the SBO transients is expected to contribute to making strategy for performing a thermal-hydraulic integral effect test with ATLAS which includes determination of proper test conditions and improvement of measurements. The primary safety criterion (PSC) was determined to be the core mixture level from the expert panel discussion. As for the major components of the nuclear power plant, the rank of the importance and the knowledge level were summarized for each of the three temporal phases. Taking into account the important phenomena from the PIRT result, it can be confirmed that ATLAS has a capability of simulating the major thermal-hydraulic phenomena during a SBO transient as realistically as possible. The present PIRT will be directly used to make strategy for performing a thermal-hydraulic integral effect test with ATLAS in the framework of the OECD-ATLAS project.

APR+ (Advanced Power Reactor Plus), which is a GEN-III+ nuclear power plant developed in Korea, adopts PAFS (passive auxiliary feedwater system) as an advanced safety feature. The PAFS can completely replace an active auxiliary feedwater system by cooling down the secondary side of steam generators with a natural convection mechanism. This study focuses on experimental and analytical investigation for cooling and operational performance of the PAFS during an FLB (feedwater line break) transient with an integral effect test facility, ATLAS-PAFS. To realistically simulate the FLB accident of the APR+, the three-level scaling methodology was taken into account to design the test facility and determine the test condition. From the test result, the PAFS was actuated to successfully cool down the decay heat of the reactor core by the condensation heat transfer at the PCHX (passive condensation heat exchanger), and thus it could be confirmed that the APR+ has the capability of coping with a FLB scenario by adopting the PAFS and proper set-points for its operation. This integral effect test data were used to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. The code analysis result proved that it could reasonably predict the FLB transient including the actuation of the PAFS and the natural convection.

New mechanistic bubble coalescence and break-up models considering turbulent suppression phenomena, which can possibly occur in the high liquid velocity condition of turbulent bubbly two-phase flow, are presented. The energy exchange mechanism between a turbulent eddy and interfacial structure was taken into account in the efficiency terms. Numerical simulations of turbulent bubbly flow were conducted in a CFD code to evaluate the newly developed models, in comparison with other advanced models coupled with a bubble-induced turbulent effect for one-group interfacial area transport equation. Local measurements of the bubble characteristics on the bubble size evolution along a vertical pipe flow were performed at KAERI-VAWL test facility using the five-sensor conductivity probe method to provide database for models validation. Results from the calculation clearly show the improvements of the newly developed models.

Based on the significance of a computational fluid dynamics (CFD) simulation of dispersed two-phase flow systems in many engineering fields, especially in nuclear energy systems, extensive analytical and numerical studies of wall lubrication force for vertical upward turbulent bubbly two-phase flow are presented. An analytical expression of the wall lubrication force considering the effect of liquid velocity is developed and analyzed. A numerical methodology for evaluating the wall lubrication force coefficients is proposed, which is based on the assumption that the balance of the forces acting on a bubble perpendicularly to the flow direction determines the establishment of radial void fraction profiles. Wall lubrication force coefficients are determined based on a reasonable agreement in radial void fraction profiles between numerical results predicted by the CFD EAGLE (elaborated analysis of gas–liquid flows evolution) code and extensive experimental database in the open literature as well as experimen...

The Passive Auxiliary Feedwater System (PAFS) is one of the advanced safety features adopted in the Advanced Power Reactor Plus (APR+). PAFS is designed to replace a conventional active Auxiliary Feedwater System (AFWS). The PAFS cools down the steam generator secondary side and eventually removes the decay heat from the reactor core by a natural circulation mechanism, i.e., condensing steam in nearly horizontal U-tubes submerged inside a pool. A separate effect test facility was constructed with the aim of validating the cooling and operational performance of the PAFS. The PAFS Condensing Heat Removal Assessment Loop (PASCAL) was constructed by simulating a single Passive Condensation Heat Exchanger (PCHX) tube submerged in the Passive Condensation Cooling Tank (PCCT) according to the volumetric scaling methodology. Quasi-steady state (SS) test cases and PCCT level decrease (PL) were sequentially performed with the steam generator heater power set at 540 kW to investigate the thermal-hydraulic behavior of the PAFS system and the characteristics of the natural circulation in the loop. The experimental results proved that the current PCHX design satisfied the heat removal requirement for cooling down the reactor core during an accident condition. Therefore, the PAFS can replace a conventional active AFWS in the APR+ by utilizing the two-phase natural circulation flow. The Multi-dimensional Analysis of Reactor Safety, KINS Standard Version (MARS-KS), a thermal hydraulic system analysis code, was utilized to validate the present experimental data. The results of the MARS-KS code provided a conservative prediction of the heat transfer phenomena for the PCHX cooling performance.

The passive auxiliary feedwater system (PAFS) is one of the advanced safety features adopted in the Advanced Power Reactor Plus (APR+) and is designed to completely replace a conventional, active a...

In the present work, two common modelling concepts taking into account the influence of bubbles on the turbulence of the liquid phase with available models in open literature were implemented in the EAGLE (Elaborated Analysis of Gas-Liquid flow Evolution) code and assessed against experimental data of Hibiki et al. (2001). The EAGLE code has been developed at KAERI based on the two-fluid model for a multi-dimensional analysis of two-phase flow with the implementation of non-drag force, standard k-ε turbulence model, and the interfacial area transport equation. In order to investigate the bubble size effect on two-phase flow evolution and to provide a data set for developing the physical models to describe the bubble-induced turbulence effect and also for validating the EAGLE code, a series of local parameter measurements as well as visualization tests were conducted in air-water vertical-upward flow condition in which the initial bubble size is controlled by a specially designed bubble generator. The numerical results and experimental data were compared, analysed and discussed in this paper.

The passive auxiliary feedwater system (PAFS) is one of the advanced safety features adopted in the APR+, which is intended to completely replace the conventional active auxiliary feedwater system. With an aim of validating the cooling and operational performance of PAFS, an experimental program is in progress at KAERI, which is composed of two kinds of tests; the separate effect test and the integral effect test. The separate effect test, PASCAL (PAFS Condensing Heat Removal Assessment Loop), is being performed to experimentally investigate the condensation heat transfer and natural convection phenomena in PAFS. A single, nearly-horizontal U-tube, whose dimensions are the same as the prototypic U-tube of the APR+ PAFS, is simulated in the PASCAL test. The PASCAL experimental result showed that the present design of PAFS satisfied the heat removal requirement for cooling down the reactor core during the anticipated accident transients. The integral effect test is in progress to confirm the operational performance of PAFS, coupled with the reactor coolant systems using the ATLAS facility. As the first integral effect test, an FLB (feedwater line break) accident was simulated for the APR+. From the integral effect test result, it could be concluded that the APR+ has the capability of coping with the hypothetical FLB accident by adopting PAFS and proper set-points of its operation.

The APR+ (Advanced Power Reactor Plus), a next generation nuclear power plant in Korea, has adopted the PAFS (Passive Auxiliary Feedwater System) on the secondary system of the steam generator (SG) as an advanced safety feature. It is intended to replace the conventional auxiliary feedwater system, which consists of active components for the SG in a passive way. It removes decay heat from the reactor core by cooling down the secondary system of the SG using a condensation heat exchanger installed in the PCCT (Passive Condensation Cooling Tank). The objective of this study is to design a condensation heat exchanger for the PAFS and to evaluate the cooling performance for the proposed design using the thermal hydraulic system analysis code, MARS (Multi-dimensional Analysis for Reactor Safety). Requirements such as the heat removal capacity and the prevention of water hammer were preferentially considered to determine the design parameters of the heat exchanger tube. The MARS code analysis result showed that the proposed design of the PAFS heat exchanger is able to cool down the required amount of decay heat. The distribution of a liquid volume fraction and flow regime predicted by the MARS code shows that the proposed design of the heat exchanger excludes the water hammer inside the tube. Estimation of a two-phase flow pressure drop indicates that the pressure drop inside the tube is negligible compared to the total pressure drop in the PAFS. From the MARS code analysis, it is concluded that the proposed design of the condensation heat exchanger in the PAFS satisfies the overall criteria for the performance of the passive heat removal system in APR+.

As passive safety features for nuclear power plants receive increasing attention, various studies have been conducted to develop safety systems for third-generation (GEN-III) nuclear power plants that are driven by passive systems, such as natural circulation, gravity, and resistance to high temperatures. Thus, South Korea has designed the Advanced Power Reactor Plus (APR+) with a two-loop PWR and 1500 MW e by adding passive safety features to the Advanced Power Reactor 1400 MW e (APR1400). The Passive Auxiliary Feedwater System (PAFS) is one of several passive safety systems being designed for the APR+, and extensive studies are being conducted to complete its design and to verify its feasibility. Because the PAFS removes decay heat from the reactor core under transient and accident conditions, it is necessary to evaluate the heat removal capability of the PAFS under hypothetical accident conditions. Therefore, in this paper, after introducing the characteristics of the PAFS and its design requirements, a performance analysis of the PAFS is performed for two accident cases: Loss of Condenser Vacuum (LOCV) and Feedwater Line Break (FLB). For the analysis, the Multi-dimensional Analysis of Reactor Safety (MARS) code is used, and a MARS model is developed by adding the PAFS model to the existing APR1400 model. The analysis results show that the PAFS has enough capacity to remove decay heat under the postulated accident conditions. In addition, the adequacy of modified control logic for main steam isolation valve (MSIV) is validated by comparing the traditional control logic.

As a series of subcooling boiling flow tests, local two-phase flow parameters were obtained at SUBO (subcooled boiling) test facility under steam–water flow conditions. The test section is a vertical annulus of which the axial length is 4.165 m with a heater rod at the center of a channel. The inner and outer diameters of the test section and the heater rod are 35.5 mm and 9.98 mm, respectively. The test was performed by a two-stage approach. Stage-I for the measurement of local bubble parameters has been already done ( Yun et al., 2009 ). The present work focused on the stage-II test for the measurement of local liquid parameters such as a local liquid velocity and a liquid temperature for a given flow condition of stage-I. A total of six test cases were chosen by following the test matrix of stage-I. The flow conditions are in the range of the heat flux of 370–563 kW/m 2 , mass flux of 1110–2100 kg/(m 2 s) and inlet subcooling of 19–31 °C at pressure condition of 0.15–0.2 MPa. From the test, local liquid parameters were measured at 6 elevations along the test section and 11 radial locations of each elevation in addition to the previously obtained local void fraction, interfacial area concentration, Sauter mean diameter and bubble velocity. The present subcooled boiling (SUBO) data completes a data set for use as a benchmark, validation and model development of the Computational Fluid Dynamics (CFD) codes or existing safety analysis codes.

To enhance the multi-dimensional analysis capability for a subcooled boiling two-phase flow, the one-group interfacial area transport equation was improved with a source term for the bubble lift-off. It included the bubble lift-off diameter model and the lift-off frequency reduction factor model. The bubble lift-off diameter model took into account the bubble's sliding on a heated wall after its departure from a nucleate site, and the lift-off frequency reduction factor was derived by considering the coalescences of the sliding bubbles. To implement the model, EAGLE (elaborated analysis of gas–liquid evolution) code was developed for a multi-dimensional analysis of two-phase flow. The developed model and EAGLE code were validated with the experimental data of SUBO (subcooled boiling) and SNU (Seoul National University) test, where the subcooled boiling phenomena in a vertical annulus channel were observed. Locally measured two-phase flow parameters included a void fraction, interfacial area concentration, and bubble velocity. The results of the computational analysis revealed that the interfacial area transport equation with the bubble lift-off model showed a good agreement with the experimental results of SUBO and SNU. It demonstrates that the source term for the wall nucleation by considering a bubble sliding and lift-off mechanism enhanced the prediction capability for the multi-dimensional behavior of void fraction or interfacial area concentration in the subcooled boiling flow. From the point of view of the bubble velocity, the modeling of an increased turbulence induced by boiling bubbles at the heated wall enhanced the prediction capability of the code.

In order to analyze thermal hydraulic phenomena during a DVI (Direct Vessel Injection) line break LOCA (Loss-of-Coolant Accident) in the APR1400 (Advanced Power Reactor 1400 MWe), we performed experimental studies with the SNUF (Seoul National University Facility), a reduced-height and reduce-pressure integral test loop with a scaled down APR1400. We performed experiments dealing with eight test cases under varied tests. As a result of the experiment, the primary system pressure, the coolant temperature, and the occurrence time of the downcomer seal clearing were affected significantly by the thermal power in the core and the SI flow rate. The break area played a dominant role in the vent of the steam. For our analytical investigation, we used the MARS code for simulation of the experiments to validate the calculation capability of the code. The results of the analysis showed good and sufficient agreement with the results of the experiment. However, the analysis revealed a weak capability in predicting the bypass flow of the SI water toward the broken DVI line, and it was insufficient to simulate the streamline contraction in the broken side. We, hence, need to improve the MARS code.

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.

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.

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.

As a result of experiments with the Upper Plenum Test Facility and the 1400-MW(electric) Advanced Power Reactor (APR1400), sweepout in the downcomer has been identified as playing an important role...

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.

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.

This study concerns the development of an integrated calculation methodology with which to continually and consistently analyze the progression of an accident from the design-basis accident phase via core uncovery to the severe accident phase. The depletion rate of reactor coolant inventory was experimentally investigated after the safety injection failure during a large-break loss-of-coolant accident utilizing the Seoul National University Integral Test Facility (SNUF), which is scaled down to 1/6.4 in length and 1/178 in area from the APR1400 [Advanced Power Reactor 1400 MW(electric)]. The experimental results showed that the core coolant inventory decreased five times faster before than after the extinction of sweepout in the reactor downcomer, which is induced by the incoming steam from the intact cold legs. The sweepout occurred on top of the spillover from the downcomer region and expedited depletion of the core coolant inventory. The test result was simulated with the MAAP4 severe accident analysis code. The calculation results of the original MAAP4 deviated from the test data in terms of coolant inventory distribution in the test vessel. After the calculation algorithm of coolant level distribution was improved by including the subroutine of pseudo pressure buildup, which accounts for the differential pressure between themore » core and downcomer in MAAP4, the core melt progression was delayed by hundreds of seconds, and the code prediction was in reasonable agreement with the overall behavior of the SNUF experiment.« less

Station blackout (SBO) accident is considered as one of the most significant design extension conditions (DECs), which has been extensively focused after the Fukushima Dai-chi accident. When the SBO accident occurs in the APR+ (Advance Power Reactor Plus), the PAFS (Passive Auxiliary Feedwater System), which is an advanced safety feature adopted in the APR+, should play a significant role to cool down the core decay heat without any operation of active safety systems. This study focuses on validation of the cooling and operational performance for the PAFS during the SBO transient with utilizing an integral effect test facility, ATLAS-PAFS. In order to simulate the SBO transient of the APR+ as realistically as possible, a pertinent scaling approach was taken into account. The initial steady-state conditions and the sequence of event in the SBO scenario for the APR+ were successfully simulated with the ATLAS-PAFS facility. In the transient simulation, major thermal-hydraulic parameters such as the system pressures, the collapsed water levels, the break flow rate, and the condensate flow rate at the return-water line were measured and investigated. Following the reactor trip at the initiation of the transient, the coolant inventory of the secondary system of the steam generator was reduced by the repeated opening and closing of the MSSV. When the collapsed water level reached 25% of wide range, the PAFS was actuated to cool down the primary system by the condensation heat transfer at the PCHX (Passive Condensation Heat Exchanger). The pressure and the temperature of the reactor coolant system continuously decreased during the heat removal by the PAFS operation. It points out that the PAFS can supply auxiliary feedwater to the steam generator and remove the core decay heat without any active system. From the present experimental result, it could be concluded that the APR+ has the capability of coping with the hypothetical SBO scenario with adopting the PAFS and proper set-points of its operation. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS, RELAP5 as well as the SPACE code and to identify any code deficiency for a SBO simulation with an operation of the PAFS.

Korea Atomic Energy Research Institute (KAERI) has designed and constructed a test facility for reactor coolant pumps (RCPs). The RCP Test Facility (RCPTF) has the capability to test a RCP under the operation condition of an Advanced Power Reactor 1400 MW (APR1400). The design values of the facility are 17.2 MPa, 343 °C, 11.7 m3/s, and 13 MW in maximum pressure, temperature, flow rate, and electrical power, respectively. In the facility, it is possible to perform a type test for a newly-developed RCP as well as a production test for a RCP before its installation in a nuclear power plant. For the production test, H-Q curves under the cold and hot conditions are acquired. For the type test, various transient tests are additionally performed including four types of seal transient tests, a thrust bearing transient test, a cost down test, and so on. To acquire H-Q curves of a RCP, the flow rate should be controlled by varying the flow resistance in the test loop. The RCPTF uses a Variable Restriction Orifice (VRO) whose flow area can be controlled by moving the two orifice plates installed in-parallel. The need for flow control valves and bypass lines was eliminated using the VRO such that the flow disturbance was minimized. The flow rate in the main loop of the RCPTF is measured by a standard venture flow meter. The flow rate in the RCPTF is very high and thus the venture flow meter could not be calibrated in the entire range of Reynolds number corresponding to the operating condition in the APR1400. The calibration was conducted at the Colorado Experiment Engineering Station Inc. (CEESI) in the USA where natural gas is used for a working fluid. If a discharge coefficient calibrated with the gas is applied in the test results performed using the water as a working fluid, a discrepancy can occur due to the static hole error. Therefore, the static hole error was compensated in the test results and the result shows the improvement. The effect of the temperature on the pressure pulsation amplitude was also evaluated. During a cold performance test and heat-up phase to the condition of a hot performance test, an abnormal increase in the pressure pulsation amplitude was observed near the specific temperature range. This is acoustic resonance phenomena that occur when a blade passing frequency of the RCP is proportional to the harmonic resonance frequency of the RCPTF.

PAFS (Passive Auxiliary Feedwater System) is one of the advanced passive safety systems adopted in the APR+ (Advanced Power Reactor plus) being developed in Korea. The PAFS aims at completely replacing the conventional active auxiliary feedwater system by cooling down the steam generator’s secondary side with a natural driving force mechanism; i.e., it can remove the core decay heat by condensing steam in a passive condensation heat exchanger (PCHX) submerged inside the passive condensation cooling tank (PCCT). With an aim of validating the cooling and operational performance of the PAFS, the separate effect test facility named as PASCAL (PAFS Condensing Heat Removal Assessment Loop) has been constructed. In this study, postulated transient scenarios occurring in the PAFS were simulated to evaluate the performance of the condensation heat transfer and investigate the thermal hydraulic phenomena of the two-phase natural convection flow. The transient test matrix is composed of inadvertent MSSV (Main Steam Safety Valve) opening test (MO), PAFS start-up actuation test (SU), and non-condensable gas effect test (NC). In the MSSV opening test, MSSV was intentionally opened and closed several times and the characteristics of the natural convection flow were investigated. The experimental results showed that the cooling performance of the PAFS could be recovered after inadvertent opening and closing the MSSV. Start-up actuation test simulated the initial transient when the PAFS actuation signal was generated and the natural convection flow was initiated in the loop, and any significant two-phase flow instability was not observed in the test. The purpose of the non-condensable gas effect test is to study the characteristics of the condensation heat transfer in the heat exchanger when the nitrogen gas was injected. The test results proved that the existence of the non-condensable gas up to 1% of the steam did not produce a meaningful decrease of the cooling capability in the PAFS. From the experimental results described above, the cooling and operating performance of the PAFS was validated with respect to occurrence of the various transient scenarios and it was proved that the function of the PAFS can be effectively performed during the transient situation. The result will be also utilized in validation of the thermal hydraulic system code in the future.

The PAFS is one of the advanced safety features adopted in the APR+ (Advanced Power Reactor Plus) which is intended to completely replace a conventional active auxiliary feedwater system. The PAFS cools down the steam generator secondary side and eventually removes the decay heat from the reactor core by adopting a natural convection mechanism; i.e., condensing steam in nearly-horizontal U-tubes submerged inside the PCCT (Passive Condensation Cooling Tank). With an aim of verifying the operational performance of the PAFS, the experimental program of an integral effect test is in progress at KAERI (Korea Atomic Energy Research Institute). The test facility, ATLAS-PAFS was constructed to experimentally investigate the thermal hydraulic behavior in the primary and secondary systems of the APR+ during a transient when the PAFS is actuated. Since the ATLAS-PAFS facility simulates a single train of the PAFS, the anticipated accident scenarios in the experiment include FLB (Feedwater Line Break), MSLB (Main Steam Line Break), and SGTR (Steam Generator Tube Rupture). Among them, SGTR was considered as one of the design basis accidents having a significant impact on safety in a viewpoint of radiological release. Therefore, the SGTR test was determined to be the integral effect test item in the frame of the ATLAS-PAFS experimental program. In this study, the PAFS-SGTR-HL-02 test was performed to simulate a double-ended rupture of a single U-tube in the hot side of the steam generator of the APR+. The three-level scaling methodology was taken into account to determine the test conditions of the steady-state and the transient. The pressures and temperatures of the system and the data related to the PAFS operation were collected with the measurement of the break flow. The initial steady-state conditions and the sequence of event of SGTR scenario for the APR+ were successfully simulated with the ATLAS-PAFS facility. And it was shown that the pressure and the temperature of the primary system were continuously decreased during the heat removal by the PAFS operation. The water pool in the PCCT was heated up to the saturation condition and the evaporation of the water made a decrease of the PCCT water level. It could be concluded from the present experimental result that the APR+ has the capability of coping with the hypothetical SGTR scenario with adopting the PAFS and the proper set-points of its operation.

The Passive Auxiliary Feedwater System (PAFS) is one of the advanced safety features adopted in the APR+ (Advanced Power Reactor Plus) which is intended to completely replace the conventional active auxiliary feedwater system. It removes the decay heat by cooling down the secondary system of the SG using condensation heat exchanger installed in the Passive Condensation Cooling Tank (PCCT). With an aim of validating the cooling and operational performance of the PAFS, PASCAL (PAFS Condensing Heat Removal Assessment Loop), was constructed to experimentally investigate the condensation heat transfer and natural convection phenomena in the PAFS. It simulates a single tube of the passive condensation heat exchangers, a steam-supply line, a return-water line, and a PCCT with a reduced area, which is equivalent to 1/240 of the prototype according to a volumetric scaling methodology with a full height. The objective of the experiment is to investigate the cooling performance and natural circulation characteristics of the PAFS by simulating a steady state condition of the thermal power. From the experiment, two-phase flow phenomena in the horizontal heat exchanger and PCCT were investigated and the cooling capability of the condensation heat exchanger was validated. Test results showed that the design of the condensation heat exchanger in PAFS could satisfy the requirement for heat removal rate of 540 kW per a single tube and the prevention of water hammer phenomenon inside the tube. It also proved that the operation of PAFS played an important role in cooling down the decay heat by natural convection without any active system. The present experimental results will contribute to improve the model of the condensation and boiling heat transfer, and also to provide the benchmark data for validating the calculation performance of a thermal hydraulic system analysis code with respect to the PAFS.