Passive Auxiliary Feedwater System Research Articles

Validation of TRACE 5.0 for steam line break with passive auxiliary feedwater system via the ATLAS facility

The TRAC/RELAP Advanced Computational Engine (TRACE) reactor systems code was validated for the Passive Auxiliary Feedwater System (PAFS) using the experimental data from the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS) facility. The accident scenario which emulates a Double-Ended Guillotine Break (DEGB) in the main steam line of Steam Generator 1, was meticulously analyzed, with a focus on transient events and steady-state conditions with an error margin of less than 2%. Comprehensive comparative analyses of key thermal hydraulic parameters, including system pressures, temperatures, flow rates, and water levels, were conducted. Discrepancies were systematically identified, investigated, and effectively resolved. The experimental results underscore the successful role of passive safety systems in mitigating potential accident impacts, emphasizing their pivotal contribution to the overall safety and reliability of nuclear reactor systems. Furthermore, the simulation results effectively mirror the experimental data, demonstrating TRACE's capability to adequately simulate the behavior of a passive feedwater system during a main steam line break.

Indirect validation of fluid-to-fluid scaling criteria for modeling steam condensation based on empirical correlations and CFD simulations

Several nuclear reactor designs have adopted passive safety systems with condensation inside horizontal tubes, such as the Passive Auxiliary Feedwater System (PAFS) of the APR+, the new generation of the Korean-Standard Pressurized Water Reactors (PWR). Performing steam condensation experiments under high-pressure conditions is needed for the design and the realistic prediction of the condensation phenomena in such systems. To avoid the costs and technical complexities associated with experimentation at high-pressure conditions, a fluid-to-fluid scaling method to model high-pressure steam condensation with a simulant fluid, refrigerant for instance, had been developed. It was determined that the liquid-to-vapor density ratio, the Froude number, and the vapor Reynolds number should be preserved in order to guarantee the flow regime and heat transfer similarity between prototypes and models. The scope of the current work is to validate the developed scaling criteria. Due to the limitation of test data of different fluids with comparable test parameters, indirect validation is proposed based on two approaches. The first approach is based on empirical correlations, in which test data from one fluid are transferred to the equivalent conditions of the other fluid, the transferred data are then compared with condensation correlations to check the validity of the scaling method. The second approach is based on CFD calculations. A CFD model, using the commercial CFD code STAR-CCM+, for the simulation of condensation in hotizontal tubes is developed. The Fluid Film model is used for modeling the condensate film thickness and development. For modeling the phase change process, source and sink terms are introduced to the continuity, momentum, and energy equations at the interface between the vapor and the condensate film. The developed CFD model is then applied in the scaling criteria validation by using it to simulate the condensation phenomena of different models in one fluid and compare its results with the corresponding experimental data in other fluid. The validation results show good agreement between prototypes and their equivalent models, and show good feasibility and reasonable accuracy of the proposed scaling criteria.

Risk reduction design alternatives for low-power and shutdown operation of a 1500-MW advanced power reactor with a passive auxiliary feedwater system

The Korean nuclear power regulatory body issued the standard design approval for a 1500-MW advanced power reactor in 2014. The reactor was designed based on a four-train independent safety concept with a passive auxiliary feedwater system. Its full power risk, or core damage frequency (CDF), was determined to be significantly lower than that of the Advanced Power Reactor 1400 (APR1400) used in Units 3 and 4 of the Shin-Kori Nuclear Power Plant. However, because the ratio of the CDF contribution to the internal full power to that of the low-power and shutdown (LPSD) is 17:83, the risk during LPSD operation requires further assessment and reduction. In this study, we investigated design alternatives for reducing LPSD operation risks and analyzed the associated sensitivities. We propose various design alternatives, and their combinations, which are analyzed based on the reduction effect they have on the CDF for plant operating states and on initiating events during LPSD operation. The proposed alternatives, and their combinations in particular, exhibited significant reductions in LPSD CDF, indicating high potential applicability.

Risk-informed approach for design optimization during low power and shutdown operation

APR+ has been developed in Korea since 2007. APR+ adopts various advanced safety features including passive auxiliary feedwater system, four emergency diesel generators. Through the implementation of the advanced designs, APR+ increased the safety to the world best level of evolutionary reactors. The full power core damage frequency or containment failure frequency decreased significantly comparing to APR1400 that is base model of APR+. However, low-power shutdown risk has not been improved substantially. This paper suggests several design changes that optimize low-power shutdown risk. Based on the design alternatives, this paper discusses risk effectiveness of the proposed design including various factors, e.g. equipment reliability, human error, training, procedure and so on.

Thermal-hydraulic study of air-cooled passive decay heat removal system for APR+ under extended station blackout

The air-cooled passive decay heat removal system (APDHR) was proposed to provide the ultimate heat sink for non-LOCA accidents. The APDHR is a modified one of Passive Auxiliary Feed-water system (PAFS) installed in APR+. The PAFS has a heat exchanger in the Passive Condensate Cooling Tank (PCCT) and can remove decay heat for 8 h. After that, the heat transfer rate through the PAFS drastically decreases because the heat transfer condition changes from water to air. The APDHR with a vertical heat exchanger in PCCT will be able to remove the decay heat by air if it has sufficient natural convection in PCCT. We conducted the thermal-hydraulic simulation by the MARS code to investigate the behavior of the APR + selected as a reference plant for the simulation. The simulation contains two phases based on water depletion: the early phase and the late phase. In the early phase, the volume of water in PCCT was determined to avoid the water depletion in three days after shutdown. In the late phase, when the number of the HXs is greater than 4089 per PCCT, the MARS simulation confirmed the long-term cooling by air is possible under extended Station Blackout (SBO).

소형화된 피동응축용 냉각탱크 내 온도분포와 비등현상에 관한 실험적 연구

An experimental study was conducted to investigate the water temperature distribution and saturated boiling phenomenon in a small-scale passive condensate cooling tank that is part of an emergency passive auxiliary feedwater system used in nuclear power plants. During the heating process using an electric heater inserted in the tank, the temperature of water was measured at 36 locations. Isothermal line contours were generated based on the experimental data. The contours indicated that the water above the heater location had a nearly uniform temperature due to free convection flow. Conversely, a strong temperature gradient was observed in the water below the heater location. The bubble departure diameters at the cylindrical heater surface were measured using images captured by a high-speed camera. The diameters were observed to be within the range of values predicted using existing correlations.

The concept of the innovative power reactor

Abstract The Fukushima accident reveals the vulnerability of existing active nuclear power plant (NPP) design against prolonged loss of external electricity events. The passive safety system is considered an attractive alternative to cope with this kind of disaster. Also, the passive safety system enhances both the safety and the economics of NPPs. The adoption of a passive safety system reduces the number of active components and can minimize the construction cost of NPPs. In this paper, reflecting on the experience during the development of the APR+ design in Korea, we propose the concept of an innovative Power Reactor (iPower), which is a kind of passive NPP, to enhance safety in a revolutionary manner. The ultimate goal of iPower is to confirm the feasibility of practically eliminating radioactive material release to the environment in all accident conditions. The representative safety grade passive system includes a passive emergency core cooling system, a passive containment cooling system, and a passive auxiliary feedwater system. Preliminary analysis results show that these concepts are feasible with respect to preventing and/or mitigating the consequences of design base accidents and severe accidents.

Performance verification of Passive Accident Mitigation Scheme (PAMS) under feedwater line break (FLB) accident

An innovative secondary circuit passive residual heat removal scheme, termed as Passive Accident Mitigation Scheme (PAMS), is investigated in our work. PAMS aims at removing core residual heat and ensuring reactor safety for Pressurized Water Reactor (PWR) Nuclear Power Plant (NPP), which consists of two parts: the first part is Passive Auxiliary Feedwater System (PAFS), and the other part is Passive Residual Heat Removal System (PRHRS).Shi et al. (2015) analyzed the PAMS operation characteristics under Station Blackout (SBO) accident using RELAP5. In order to assess the ability of PAMS, more analyses should be carried out. This paper takes the Westinghouse-designed Advanced Passive PWR (AP1000) as research object and proposes the simulation performance verification of PAMS under FLB accident with the help of RELAP5. Moreover, the comparative analysis is also conducted between PAMS and traditional secondary circuit PRHRS with different design capacities under FLB accident. Results show that: during the whole accident process, the key parameters’ performances of primary system could meet acceptance criteria requirements, which proves PAMS can cope with FLB accident; moreover, since only the PAMS train connected to intact circuit can operate to mitigate FLB accident, the analysis results prove that PAMS design satisfies single failure criterion; the primary system cooling process with PAMS operation is quite close to primary system cooling process with 4% of Full Power (FP) design capacity PRHRS operation, and both of them can remove core residual heat significantly under FLB accident.

Evaluation of mechanistic wall condensation models for horizontal heat exchanger in PAFS (Passive Auxiliary Feedwater System)

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.

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.

Prediction of nucleate boiling heat transfer on horizontal U-shaped heat exchanger submerged in a pool of water using MARS code

In advanced nuclear power plants, a horizontal U-shaped heat exchanger submerged in a pool is under development as a key equipment of a passive safety system. For the successful design of the heat exchanger and the safety analysis of the nuclear power plant incorporating this passive safety system, the reliable prediction of the nucleate boiling heat transfer on the horizontal parts of the U-shaped tubes is one of the important factors. At present, the best estimate thermal hydraulic analysis codes such as RELAP5 and MARS are used to analyze the nucleate boiling heat transfer on the horizontal U-shaped heat exchanger submerged in a pool; however, it is still not known how to physically model the heat exchanger pool, and which correlations are suitable among the pool boiling and forced convective boiling correlations. To secure the applicable correlation for the heat exchanger, this study assessed 15 nucleate boiling correlations using MARS. To improve the prediction capability of the best estimate code, this study investigated the main heat transfer mechanisms on the horizontal U-shaped heat exchanger submerged in a pool, then proposed a prediction method, and finally developed a nucleate boiling model. From the validation of the proposed model against PAFS (passive auxiliary feedwater system)-related experimental data of PASCAL and ATLAS-PAFS, the proposed boiling model predicted the experimental heat transfer coefficients much better than the default nucleate boiling model by Chen (1966, Ind. Eng. Chem. Process Des. Dev. 5 (3), 329) in MARS. A total of 366 data points were predicted with a mean deviation of 8.1%. Hence, the proposed nucleate boiling model appears to be usable to the reliable design and the safety analysis of the passive safety system using the best estimate codes.

Parametric experiments and CFD analysis on condensation heat transfer performance of externally condensing tubes

To ensure safe operation of nuclear power plants even in the case of a prolonged station blackout, advanced reactors adopt passive systems that can operate without electricity supply. In Korea, a passive auxiliary feedwater system was successfully validated, and a passive containment cooling system (PCCS) has recently attracted attention. To investigate the thermal performance of PCCSs, we perform various experiments with external heat exchangers, which condense steam externally, for PCCSs. Through experiments, we construct a database for the lower air mass fraction and perform a parametric study on the tube inclination and diameter. The operating ranges for the experiments are 0.24–0.38MPa (pressure), 0.06–0.4 (air mass fraction), and 0–90° (tube inclination). A lower tube inclination and smaller tube diameter are found to yield higher heat-transfer coefficients, by approximately 20%. In the prediction of condensation heat-transfer coefficients, experimental correlations and the heat–mass transfer analogy have limitations in both accuracy and applicability. A computational-fluid-dynamics approach is used with the aid of user-defined functions to calculate the heat-transfer coefficients. The resulting predictions exhibit an average error of 7% when the air mass fraction is higher than 0.2.

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.

The investigation of Passive Accident Mitigation Scheme for advanced PWR NPP

To enhance inherent safety features of nuclear power plant, the advanced pressurized water reactors implement a series of passive safety systems. This paper puts forward and designs a new Passive Accident Mitigation Scheme (PAMS) to remove residual heat, which consists of two parts: the first part is Passive Auxiliary Feedwater System (PAFS), and the other part is Passive Heat Removal System (PHRS).This paper takes the Westinghouse-designed Advanced Passive PWR (AP1000) as research object and analyzes the operation characteristics of PAMS to cope with the Station Blackout Accident (SBO) by using RELAP5 code. Moreover, the comparative analysis is also conducted between PAMS and Traditional Secondary Circuit PHRS to derive the advantages of PAMS. The results show that the designed scheme can remove core residual heat significantly and maintain the plant in safe conditions; the first part of PAMS would stop after 120min and the second part has to come into use simultaneously; the low pressurizer (PZR) pressure signal would be generated 109min later caused by coolant volume shrinkage, which would actuate the Passive Safety Injection System (PSIS) to recovery the water level of pressurizer; the flow instability phenomenon would occur and last 21min after the PHRS start-up; according to the comparative analysis, the coolant average temperature gradient and the Passive Condensate Cooling Tank (PCCT) water temperature rising amplitude of PAMS are lower than those of Traditional Secondary Circuit PHRS.

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.

Experimental investigation on the natural convection flow in pool boiling

In the present study, the key thermal hydraulic phenomena within a passive condensate cooling tank (PCCT) of a small-scale pool test rig with a single heater rod are experimentally investigated. The volumetric scaling ratio of the test rig is 1/910 the size of the passive auxiliary feedwater system (PAFS) condensing heat removal assessment loop (PASCAL), which is a PAFS performance evaluation test facility. The two-dimensional velocity vector fields that occur as the water level decreases are experimentally investigated in a pool that contains a horizontal heater rod. The 2D particle image velocimetry (PIV) measurement technique is adopted to determine the velocity vector field of the natural convection flow. The experimental results indicate that a large natural convection flow occurs above the heater rod and that thermal stratification occurs below the heater rod. The thermal stratification and the stagnant region begin to disappear when the pool temperature reaches approximately 90°C. The experimental results can provide benchmark data to validate computational fluid dynamics (CFD) calculations of thermal hydraulic phenomena that occur in a pool with a heat source.

Analytical Study of Flow Instability in a Natural Circulation Pressurized Water Reactor Auxiliary Feedwater System1

A pressurized water reactor (PWR) incorporates a passive auxiliary feedwater system (PAFS), a closed natural circulation loop which is aligned to feed condensed water to its corresponding steam generator (SG). During its operation, saturated steam in the SG secondary side moves up due to buoyancy force and passes through a steam line, and then flows into a tube-tank type passive condensation heat exchanger (PCHX) where steam is condensed inside the tubes while the tube outer surfaces are cooled by the pool water. The condensate water is passively fed into the bottom of the SG secondary side by gravity. Because a natural circulation loop is susceptible to two-phase flow instability, it is requisite to confirm the system is designed adequately to avoid the potential challenges to its operational safety due to the instability. This paper presents an analytical approach for assessing if the PAFS has possible thermal and fluid mechanical characteristics which could lead to an undesirable unstable or oscillating condensate water level in the vertical pipe section. Both steady and unsteady analytical solutions for a simplified natural circulation loop model of the PAFS were derived in terms of the condensate water level and velocity in the vertical pipe section. From the solutions, the criteria for determining a potential for flow instability in the system were obtained.

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.