Power Reactor Plus Research Articles

연성회로기판 기반 수평전열관 표면의 비등기포거동 가시화 실험 연구

The Passive Auxiliary Feedwater System(PAFS) is one of the advanced safety concepts adopted in the Advanced Power Reactor Plus(APR+). To validate the operational performance of the PAFS, detailed understanding of a boiling heat transfer on horizontal tube outside is of great importance. Especially, in the mechanistic boiling heat transfer model, it is important to visualize the phenomena but there are some limitations with conventional experimental approaches. In the present study, we devised a heater based on the Flexible Printed Circuit Board (FPCB) for a more comprehensive visualization and subsequently, a digital image processing technique for the bubble motion measurement was established. Using the measurement technique, important parameters of the nucleate boiling are analyzed.

Modeling of high pressure steam condensation in inclined horizontal tubes of PAFS in APR+

ABSTRACTA Single-tube COndensation exPeriment (SCOP) was performed at Korea Advanced Institute of Science and Technology (KAIST) to simulate the condensation phenomena of Passive Auxiliary Feedwater System (PAFS), which is introduced in Advanced Power Reactor Plus. SCOP tests were conducted for two pipes of different diameters [22.6 and 44.8 mm inside diameter] with 8.2 m length and 2.5–4 degrees of the inclined angles of the upper and lower section of each tube. The experiment were performed at 1.5–6.5 MPa of steam pressure and 0.12–0.45 kg/s of steam flow. Based on the Kim and No correlation (KAIST, 2000), we developed a new turbulent–condensation correlation for the turbulent region of the liquid film in the inclined tube. We introduced the factors to account for the different inclined angle effect, different diameter effect. The Nusselt correlation is adopted for condensation near the tube inlet where the laminar region of the liquid film is dominant due to the small amount of condensed water. The present correlation was compared with the Shah and RELAP5 condensation model using the SCOP data. The root mean square errors of the present correlation, Shah model, and RELAP5 model are 17.41%, 57.88%, and 43.07%, respectively: the present correlation can predict well the condensation characteristics in comparison with other correlations in PAFS.

Comparative study for modeling reactor internal geometry in CFD simulation of PWR and PHWR internal flow – Nuclear regulatory perspective

In this study, in order to examine the effect of reactor internal geometry modeling method on the prediction accuracy for PWR (Pressurized Water Reactor) and PHWR (Pressurized Heavy Water Reactor) internal flow distribution, simulations with real geometry modeling were conducted with the commercial CFD (Computational Fluid Dynamics) software, ANSYS CFX R.14 and FLUENT R.14. The predicted results were compared with those of the porous medium assumption and the measured data. Test cases were 1/5 scaled-down APR+ (Advanced Power Reactor Plus) model for PWR and moderator test vessel (Stern Laboratories Inc.) for PHWR. It was concluded that core inlet flow distribution could be predicted more accurately by considering the real geometry of the scale-down APR + internal structures, located in the upstream of core inlet. Therefore, if the sufficient computation resource is available, an exact representation of these internal structures, for examples lower support structure bottom plate and ICI (In-Core Instrumentation) nozzle support plate, is essential for the accurate simulation of PWR internal flow. For PHWR, an approach to consider the real geometry of tubes predicted the local velocity and temperature distribution better than the Porous Medium Model. However, the benefits (e.g. improvements of prediction accuracy) against the computation cost were small; compared with PWR case. The reason may be that relatively simple PHWR internal structure can make it possible to establish a generally applicable correlation for the pressure loss coefficient.

Void fraction prediction for separated flows in the nearly horizontal tubes

Abstract A mechanistic model for void fraction prediction with improved interfacial friction factor in nearly horizontal tubes has been proposed in connection with the development of a condensation model package for the passive auxiliary feedwater system of the Korean Advanced Power Reactor Plus. The model is based on two-phase momentum balance equations to cover various types of fluids, flow conditions, and inclination angles of the flow channel in a separated flow. The void fraction is calculated without any discontinuity at flow regime transitions by considering continuous changes of the interfacial geometric characteristics and interfacial friction factors across three typical separated flows, namely stratified–smooth, stratified–wavy, and annular flows. An evaluation of the proposed model against available experimental data covering various types of fluids and flow regimes showed a satisfactory agreement.

Development of an operation strategy for a hybrid safety injection tank with an active system

Abstract A hybrid safety injection tank (H-SIT) can enhance the capability of an advanced power reactor plus (APR+) during a station black out (SBO) that is accompanied by a severe accident. It may a useful alternative to an electric motor. The operations strategy of the H-SIT has to be investigated to achieve maximum utilization of its function. In this study, the master logic diagram (i.e., an analysis for identifying the differences between an H-SIT and a safety injection pump) and an accident case classification were used to determine the parameters of the H-SIT operation. The conditions that require the use of an H-SIT were determined using a decision-making process. The proper timing for using an H-SIT was also analyzed by using the Multi-dimensional Analysis of Reactor Safety (MARS) 1.3 code (Korea Atomic Energy Research Institute, Daejeon, South Korea). The operation strategy analysis indicates that a H-SIT can mitigate five types of failure: (1) failure of the safety injection pump, (2) failure of the passive auxiliary feedwater system, (3) failure of the depressurization system, (4) failure of the shutdown cooling pump (SCP), and (5) failure of the recirculation system. The results of the MARS code demonstrate that the time allowed for recovery can be extended when using an H-SIT, compared with the same situation in which an H-SIT is not used. Based on the results, the use of an H-SIT is recommended, especially after the pilot-operated safety relief valve (POSRV) is opened.

ICONE23-2089 INTEGRAL EFFECT TEST ON OPERATIONAL PERFORMANCE OF PAFS (PASSIVE AUXILIARY FEEDWATER SYSTEM) FOR CIV OPENING STROKE

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. 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). PAFS-CIV-01 test was performed for validating cooling rate of the reactor according to the CIV opening stroke at the FLB accident, which was analyzed as the most severe case in the APR+ SSAR (Standard Safety Analysis Report). With an aim of simulating a FLB+CIV accident of the APR+ as realistically as possible, the three-level scaling methodology was taken into account to determine the test conditions of the steadystate and the transient. From the experiment, major thermalhydraulic phenomena such as the system pressures, the collapsed water levels, the break flow rate, and the condensate flow rate in PAFS were investigated and discussed. With decreasing the flow area of the CIV, the gradient of system pressure and temperature were reduced due to the decrease of the heat removal performance. As the CIV was re-opened, the two-phase natural convection flow in the loop of the PAFS was recovered. From the present experimental result, it could be concluded that the cooling rate of the core was controlled by the adjustment of the CIV opening stroke when the APR+ PAFS was operating.

Thermal-hydraulic evaluation of passive containment cooling system of improved APR+ during LOCAs

Various reactor concepts and technologies have been devised and evaluated to ensure the integrity of the core and the containment under a prolonged station blackout. After the successful validation of the passive auxiliary feedwater system (PAFS) of Advanced Power Reactor Plus (APR+), Korea is considering an improved APR+ with a passive containment cooling system (PCCS). In a previous paper, we suggested a PCCS design based on APR+ and performed a scoping analysis. We performed a MARS simulation for the thermal hydraulic evaluation of the system behavior, including natural circulation through the Passive Condensation Cooling Tank (PCCT) water pool and inside PCCS tubes as well as steam–air mixture condensation inside the containment. Through a simulation using the MARS system code, we investigated the effect of air holdup tanks (AHTs) on reduction of the air fraction in the containment, the effects of several steam–air mixture condensation models, the role of heat structures, and the flow instability inside the PCCS tubes. We found that the presence of AHT reduced the number of required PCCS tubes by more than half and the heat resistance of the steam–air mixture side is dominant in terms of governing the overall performance of PCCS. The embedded MARS condensation model and Uchida's correlation gave lower heat transfer coefficients than Dehbi's correlation, and heat structures removed more decay heat than PCCS tubes. Finally, intense flow instability inside the PCCS tubes was observed and was mitigated by placing orifice plates at the inlet of the PCCS tubes, increasing the height of the return nozzle, or increasing the tube angle.

Integral effect test and code analysis on the cooling performance of the PAFS (passive auxiliary feedwater system) during an FLB (feedwater line break) accident

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.

Heat structure coupling of CUPID and MARS for the multi-scale simulation of the passive auxiliary feedwater system

For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal hydraulics code, named CUPID, has been developed. In the present study, the CUPID code was coupled with a system analysis code MARS in order to apply it for the multi-scale thermal-hydraulic analysis of the passive auxiliary feedwater system (PAFS). The PAFS is one of the advanced safety features adopted in the Advanced Power Reactor Plus (APR+), which is intended to completely replace the conventional active auxiliary feedwater system. For verification of the coupling and validation of the coupled code, the PASCAL test facility was simulated, which was constructed with an aim of validating the cooling and operational performance of the PAFS. The two-phase flow phenomena of the steam supply system including the condensation inside the heat exchanger tube were calculated by MARS while the natural circulation and the boil-off in the large water pool that contains the heat exchanger tube were simulated by CUPID. This paper presents the description of the PASCAL facility, the coupling method and the simulation results using the coupled code.

Comparative study on the effect of reactor internal structure geometry modeling methods on the prediction accuracy for PWR internal flow distribution

Internal structures, especially those located in the upstream of a reactor core, may have a significant influence on the core-inlet flow-rate distribution; depending on both their shapes and the relative distance between the internal structures and the core inlet. In this study, in order to examine the effect of the method used to model the geometry of the internal structure of reactors, on the accuracy of prediction for the scaled-down APR+(Advanced Power Reactor Plus) flow distribution, simulations with real geometry modeling were conducted with the commercial computational fluid dynamics software, ANSYS CFX R.14. The results predicted were compared with those produced with a porous medium assumption. It was concluded that the core-inlet flow distribution could be predicted more accurately by considering the real geometry of the internal structures located in the upstream of the core inlet. Therefore, if sufficient computing power is available, an exact representation of these internal structures (e.g., the lower-support-structure bottom-plate and ICI (In-Core Instrumentation) nozzle-support plate), is needed for accurate simulation of reactor internal flow.

원자로 내부유동 예측을 위한 상용 전산유체역학 소프트웨어 성능 비교 연구

전산유체역학 소프트웨어의 일부 개발자 및 사용자는 최신 전산유체역학 소프트웨어가 최소한 단상 원자로 안전문제는 타당하게 해석할 수 있을 것으로 생각하지만 계산 결과에는 여전히 제한성 및 불확실도가 존재한다. 현재 한국원자력안전기술원에서는 규제관점에서 원자로 안전문제에 대한 상용 전산유체역학 소프트웨어의 성능평가를 수행하고 있다. 본 연구에서는 축소 APR+ 원자로 내부유동 해석시 다공성 모델을 적용한 상용 전산유체역학 소프트웨어의 예측 성능을 평가하기 위해 ANSYS CFX R.14 및 FLUENT R.14 에 탑재된 수치모델을 이용하여 계산을 수행하였다. 결론적으로 전산유체역학 소프트웨어에 따라 축소 APR+ 원자로 내부유동 분포는 국부적으로 차이가 발생하였다. 비록 제한된 수의 측정치로 인해 상용 전산유체역학 소프트웨어간 예측성능을 평가하기에는 다소 한계가 있으나 CFX R.14 가 FLUENT R.14 에 비해 상대적으로 타당한 예측결과를 제시하였다. 한편 적용된 차분법의 차이로 인해 동일한 격자에 대해 FLUENT R.14 가 CFX R.14 에 비해 상대적으로 많은 계산 메모리를 필요로 하였다. 따라서 대용량 병렬 계산시 가용한 계산 자원에 적합한 전산유체역학 소프트웨어가 선정되어야 한다.

ASSESSMENT OF CONDENSATION HEAT TRANSFER MODEL TO EVALUATE PERFORMANCE OF THE PASSIVE AUXILIARY FEEDWATER SYSTEM

As passive safety features for nuclear power plants receive increasing attention, various studies have been conducted to develop safety systems for 3rd-generation (GEN-III) nuclear power plants that are driven by passive systems. The Passive Auxiliary Feedwater System (PAFS) is one of several passive safety systems being designed for the Advanced Power Reactor Plus (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. The heat removal capability of the PAFS is strongly dependent on the heat transfer at the condensate tube in Passive Condensation Heat Exchanger (PCHX). To evaluate the model of heat transfer coefficient for condensation, the Multi-dimensional Analysis of Reactor Safety (MARS) code is used to simulate the experimental results from PAFS Condensing Heat Removal Assessment Loop (PASCAL). The Shah model, a default model for condensation heat transfer coefficient in the MARS code, under-predicts the experimental data from the PASCAL. To improve the calculation result, The Thome model and the new version of the Shah model are implemented and compared with the experimental data.

Experimental study of the APR+ reactor core flow and pressure distributions under 4-pump running conditions

The core inlet flow rates and exit pressure distributions of an APR+ (Advanced Power Reactor Plus) reactor were evaluated experimentally with the ACOP (APR+ Core Flow & Pressure) test facility. The ACOP test facility was constructed with a linear reduced scale of 1/5 referring to the APR+ reactor. The major flow path from the clod leg to hot leg was preserved with a principle of similarity. The core region was simulated using 257 core simulators, which are representative of the real HIPER fuel assemblies that APR+ reactor adopted. The core inlet flow rates and pressure distributions along the main flow path, which are significant information as an input data to evaluate the core thermal margin and reactor safety, were obtained by differential pressures measured at core simulators representing 257 fuel assemblies, and the static or differential pressures at 584 points, respectively. Two kinds of experiments, 4-pump balanced and unbalanced flow conditions, were conducted to examine the hydraulic characteristics of the reactor coolant flow. The mass balance and overall pressure drop were carefully examined to check the reliability of the obtained values. The inlet flow rates of the two test results showed similar distributions, which met the hydraulic performance requirement. The details of these experiments, the facility, and a data analysis are also described in this paper.

축소 APR+ 원자로 모형에서의 내부유동분포 수치해석

A series of 1/5 scale-down reactor flow distribution tests had been conducted to determine the hydraulic characteristics of an APR+ (Advanced Power Reactor Plus), which were used as the input data for an open core thermal margin analysis code. In this study, to examine the applicability of computational fluid dynamics with the porous model to the analysis of APR+ internal flow, simulations were conducted using the commercial multi-purpose computational fluid dynamics software ANSYS CFX V.14. It was concluded that the porous domain approach for some reactor internal structures could adequately predict the flow characteristics inside a reactor in a qualitative manner. If sufficient computational resources are available, the predicted core inlet flow distribution is expected to be more accurate by considering the real geometry of the internal structures, especially upstream of the core inlet.

A numerical study for the effect of flow skirt geometry on reactor internal flow

A series of 1/5 scale reactor flow distribution tests had been conducted in order to determine the hydraulic characteristics of the APR+ (Advanced Power Reactor Plus) which were used as the input data for the open core thermal margin analysis code. In this study, in order to examine the applicability of computational fluid dynamics with the porous model in the analysis of reactor internal flow, simulation was conducted with the commercial multi-purpose computational fluid dynamics software, ANSYS CFX V.14. In addition, among the various reactor internals the effect of flow skirt geometry on reactor internal flow was investigated. It was concluded that depending on the shape of flow skirt the flow distribution was locally somewhat different. Standard deviation of mass flow rate (σ) for the original shape of flow skirt was smaller than that for the modified shape of flow skirt. This means that the original shape of flow skirt may give the more uniform distribution of mass flow rate at core inlet plane, which may be more desirable for the core cooling. Porous model for some reactor internal structures could adequately predict the hydraulic characteristics inside reactor in a qualitative manner. However, while the predicted high core inlet flow rate region was located in the core center zone, the measured one was located in the core outer boundary. This difference may result from the fact that some internal structures including the lower support structure assembly were not modeled with the real geometry but treated as the porous domain. If the sufficient computation resource is available, the predicted core inlet mass flow distribution is expected to be more accurate by considering the real geometry of the internal structures, especially located in the upstream of core inlet.

노심손상빈도 평가를 위한 APR+ PAFS의 안전 해석

The Advanced Power Reactor Plus(APR+), which is a GEN III+ reactor based on the APR1400, is being developed in Korea. In order to enhance the safety of the APR+, a passive auxiliary feedwater system(PAFS) has been adopted in the APR+. The PAFS replaces the conventional active auxiliary feedwater system(AFWS) by introducing a natural driving force mechanism while maintaining the system function of cooling the primary side and removing the decay heat. As the PAFS completely replaces the conventional AFWS, it is required to verify the cooling capacity of PAFS for the core damage frequency(CDF) evaluation. For this reason, this paper discusses the cooling performance of the PAFS during transient accidents. The test case and scenarios were picked from the result of the sensitivity analysis in APR+ Probabilistic Safety Assessment(PSA). The analysis was performed by the best estimate thermal-hydraulic code, RELAP5/.MOD3.3. This study shows that the plant maintains the stable state without the core damages under the given test scenarios. The results of PSA considering this analysis' results shows that the CDF values are decreased. The analysis results can be used for more realistic and accurate performance of a PSA.

유동 덮개 형상이 축소 APR+ 내부 유동분포에 미치는 영향에 대한 수치해석

In this study, in order to examine the applicability of computational fluid dynamics with the porous model to the analysis of APR+ (Advanced Power Reactor Plus) internal flow, simulation was conducted with the commercial multi-purpose computational fluid dynamics software, ANSYS CFX V.14. In addition, among the various reactor internals, the effect of flow skirt geometry on reactor internal flow was investigated. It was concluded that the porous model for some reactor internal structures could adequately predict the hydraulic characteristics inside the reactor in a qualitative manner. If sufficient computation resource is available, the predicted core inlet flow distribution is expected to be more accurate, by considering the real geometry of the internal structures, especially located in the upstream of the core inlet. Finally, depending on the shape of the flow skirt, the flow distribution was somewhat different locally. The standard deviation of the mass flow rate (<TEX>${\sigma}$</TEX>) for the original shape of flow skirt was smaller, than that for the modified shape of flow skirt. This means that the original shape of the flow skirt may give a more uniform distribution of mass flow rate at the core inlet plane, which may be more desirable for the core cooling.

An experimental study on the validation of cooling capability for the Passive Auxiliary Feedwater System (PAFS) condensation heat exchanger

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 540kW 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.

Experimental Investigation into the Effect of the Passive Condensation Cooling Tank Water Level in the Thermal Performance of the Passive Auxiliary Feedwater System

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 auxiliary feedwater system. With the aim of validating the cooling and operational performance of the PAFS, a separate effect test facility, 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. During heat removal of the PAFS, the pool water in the PCCT plays a role in the ultimate heat sink of a decay heat. In this study, the effect of the PCCT water level on the cooling performance of the PAFS was experimentally investigated with the PASCAL facility. Quasi-steady-state and PCCT level decrease test cases were sequentially performed by varying the steam generator heater power from 300 to 750 kW to investigate the thermal-hydraulic behavior during the decrease of the PCCT water level. From the experimental results, it was found that the decrease of the PCCT water level enhanced evaporative heat transfer at the outer wall of the PCHX tube by reducing the degree of subcooling around the PCHX. That induced an increase of the heat removal rate by the PCHX during the transient. Thus, it can be concluded that the current design of the PCHX in the PAFS has sufficient capacity to cool down the decay heat during the whole transient of the PCCT water level decrease.

ADVANCED DVI+

A new advanced safety feature of DVI+ (Direct Vessel Injection Plus) for the APR+ (Advanced Power Reactor Plus), to mitigate the ECC (Emergency Core Cooling) bypass fraction and to prevent switching an ECC outlet to a break flow inlet during a DVI line break, is presented for an advanced DVI system. In the current DVI system, the ECC water injected into the downcomer is easily shifted to the broken cold leg by a high steam cross flow which comes from the intact cold legs during the late reflood phase of a LBLOCA (Large Break Loss Of Coolant Accident)For the new DVI+ system, an ECBD (Emergency Core Barrel Duct) is installed on the outside of a core barrel cylinder. The ECBD has a gap (From the core barrel wall to the ECBD inner wall to the radial direction) of 3/25~7/25 of the downcomer annulus gap. The DVI nozzle and the ECBD are only connected by the ECC water jet, which is called a hydrodynamic water bridge, during the ECC injection period. Otherwise these two components are disconnected from each other without any pipes inside the downcomer. The ECBD is an ECC downward isolation flow sub-channel which protects the ECC water from the high speed steam crossflow in the downcomer annulus during a LOCA event. The injected ECC water flows downward into the lower downcomer through the ECBD without a strong entrainment to a steam cross flow. The outer downcomer annulus of the ECBD is the major steam flow zone coming from the intact cold leg during a LBLOCA. During a DVI line break, the separated DVI nozzle and ECBD have the effect of preventing the level of the cooling water from being lowered in the downcomer due to an inlet-outlet reverse phenomenon at the lowest position of the outlet of the ECBD.