• Journal of Energy Engineering
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• v.24 no.3
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• pp.96-108
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• 2015

A passive containment cooling system has been designed to remove the heat inside a containment during accidents without external power supply. In this work, the PCCS was introduced in the APR1400 plant to replace the containment spray system and, then, the thermal-hydraulic performance of the PCCS was analyzed using the system thermal-hydraulic computer code, MARS. A double-ended cold-leg break accident, which is known to induce the maximum pressure in the containment, is simulated, where the thermal hydraulics of the PCCS, the reactor coolant system, and the containment are simultaneously simulated. The results of the calculations showed that the PCCS can replace the existing spray system and that the containment building and its internal structure also play a very important role for the heat removal during the accident. Some sensitivity calculations were carried out to evaluate the model uncertainty and the effects of design parameters. The limitations of the PCCS are also discussed.

• Journal of the Korean Society of Safety
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• v.35 no.3
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• pp.96-104
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• 2020

The numerical simulations were conducted to investigate the thermal-fluid phenomena occurred inside the experimental apparatus during a PCCS, used to remove heat released in accidents from a containment of light water nuclear power plant, operation. Numerical simulations of the flow and heat transfer caused by wall condensation inside the containment simulation vessel (CSV), which equipped with 18 vertical heat exchanger tubes, were conducted using the commercial computational fluid dynamics (CFD) software ANSYS-CFX. Shear stress transport (SST) and the wall condensation model were used for turbulence closure and wall condensation, respectively. The simulation using the actual size of the apparatus. However, rather than simulating the whole experimental apparatus in consideration of the experimental cases, calculation resources, and calculation time, the simulation model was prepared only in CSV. Selective simulation was conducted to verify the effects of non-condensable gas(NC gas) concentration, CSV internal pressure, and wall sub-cooling conditions. First, as a result of the internal flow of CSV, it was observed that downward flow due to condensation occurred surface of the vertical tube and upward flow occurred in the distant place. Natural convection occurred actively around the heat exchanger tube. Due to this rising and falling internal flow, natural circulation occurred actively around the heat exchanger tubes. Next, in order to check the performance of built-in condensation model using according to the non-condensable gas concentration, CSV internal flow and wall sub-cooling, the heat flux values were compared with the experimental results. On average, the results were underestimated with and error of about 25%. In addition, the influence of CSV internal pressure and wall sub-cooling was small, but when the condensate was highly generated due to the low non-condensable gas concentration, the error was large compared to the experimental values. This is considered to be due to the nature of the condensation model of the CFX code. However, in spite of the limitations of CFD, it is valid to use the built-in condensation model of CFD for PCCS performance prediction from a conservative perspective.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1125-1139
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• 2023

The system-integrated modular advanced reactor 100 (SMART100), an integral-type pressurized water small modular reactor, is based on a novel design concept for containment cooling and radioactive material reduction; it is known as the containment pressure and radioactivity suppression system (CPRSS). There is a passive cooling system using a condensation with non-condensable gas in the SMART CPRSS. When a design basis accident such as a small break loss of coolant accident (SBLOCA) occurs, the pressurized low containment area (LCA) of the SMART CPRSS leads to steam condensation in an incontainment refuelling water storage tank (IRWST). Additionally, the steam and non-condensable gas mixture passes through the CPRSS heat exchanger (CHX) submerged in the emergency cooldown tank (ECT) that can partially remove the residual heat. When the steam and non-condensable gas mixture passes through the CHX, the non-condensable gas can interrupt the condensation heat transfer in the CHX and it degrades CHX performance. In this study, condensation heat transfer experiments of steam and non-condensable gas mixture in the natural circulation loop were conducted. The pressure, temperature, and effects of the non-condensable gas were investigated according to the constant inlet steam flow rate with non-condensable gas injections in the loop.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1431-1441
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• 2017

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.

• Nuclear Engineering and Technology
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• v.27 no.3
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• pp.281-291
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• 1995

The objective of this work is to obtain the experimental data for the heat transfer processes occurring both on the inside and outside surfaces of containment steel wall with dry and wet outer surface conditions in the passive containment cooling system. The test model represented a 60$^{\circ}$ section of a containment vessel based on the AP 600 geometry. Major linear dimensions of the test model ore reduced tv a factor of ten. To simulate the decay heat a steam generator heated by electricity was placed in the test model. The maximum heat flux was 8.91 kW/$m^2$. Two types of tests were performed. The one was the tort on the natural convection of air without water film flow. The other was the evaporative heat transfer test with the falling water film flow and natural air draft. no test result shooed that the heat transfer capability by the natural convection from the containment to the air without oater film flow was limited at about 1.48 kW/$m^2$ heat flux. It was found that the heat removal capability was remarkably enhanced in the tests with the waster film flow and air draft. The obtained heat transfer data ore compared with the existing correlations.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2297-2310
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• 2022

A condensation heat transfer model is essential to accurately predict the performance of the passive containment cooling system (PCCS) during an accident in an advanced light water reactor. However, most of existing models tend to predict condensation heat transfer very well for a specific range of thermal-hydraulic conditions. In this study, a new correlation for condensation heat transfer coefficient (HTC) is presented using machine learning technique. To secure sufficient training data, a large number of pseudo data were produced by using ten existing condensation models. Then, a neural network model was developed, consisting of a fully connected layer and a convolutional neural network (CNN) algorithm, DenseNet. Based on the hold-out cross-validation, the neural network was trained and validated against the pseudo data. Thereafter, it was evaluated using the experimental data, which were not used for training. The machine learning model predicted better results than the existing models. It was also confirmed through a parametric study that the machine learning model presents continuous and physical HTCs for various thermal-hydraulic conditions. By reflecting the effects of individual variables obtained from the parametric analysis, a new correlation was proposed. It yielded better results for almost all experimental conditions than the ten existing models.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.803-811
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• 2024

Nuclear accidents such as Fukushima Daiichi have highlighted the potential of passive safety systems to replace or complement active safety systems as part of the overall prevention and/or mitigation strategies. In addition, passive systems are key features of Small Modular Reactors (SMRs), for which they are becoming almost unavoidable and are part of the basic design of many reactors available in today's nuclear market. Nevertheless, their potential to significantly increase the safety of nuclear power plants still needs to be strengthened, in particular the ability of computer codes to determine their performance and reliability in industrial applications and support the safety demonstration. The PASTELS project (September 2020-February 2024), funded by the European Commission "Euratom H2020" programme, is devoted to the study of passive systems relying on natural circulation. The project focuses on two types, namely the SAfety COndenser (SACO) for the evacuation of the core residual power and the Containment Wall Condenser (CWC) for the reduction of heat and pressure in the containment vessel in case of accident. A specific design for each of these systems is being investigated in the project. Firstly, a straight vertical pool type of SACO has been implemented on the Framatome's PKL loop at Erlangen. It represents a tube bundle type heat exchanger that transfers heat from the secondary circuit to the water pool in which it is immersed by condensing the vapour generated in the steam generator. Secondly, the project relies on the CWC installed on the PASI test loop at LUT University in Finland. This facility reproduces the thermal-hydraulic behaviour of a Passive Containment Cooling System (PCCS) mainly composed of a CWC, a heat exchanger in the containment vessel connected to a water tank at atmospheric pressure outside the vessel which represents the ultimate heat sink. Several activities are carried out within the framework of the project. Different tests are conducted on these integral test facilities to produce new and relevant experimental data allowing to better characterize the physical behaviours and the performances of these systems for various thermo-hydraulic conditions. These test programmes are simulated by different codes acting at different scales, mainly system and CFD codes. New "system/CFD" coupling approaches are also considered to evaluate their potential to benefit both from the accuracy of CFD in regions where local 3D effects are dominant and system codes whose computational speed, robustness and general level of physical validation are particularly appreciated in industrial studies. In parallel, the project includes the study of single and two-phase natural circulation loops through a bibliographical study and the simulations of the PERSEO and HERO-2 experimental facilities. After a synthetic presentation of the project and its objectives, this article provides the reader with findings related to the physical analysis of the test results obtained on the PKL and PASI installations as well an overall evaluation of the capability of the different numerical tools to simulate passive systems.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.144-155
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• 2023

The core catcher is used as a passive safety system in new generation nuclear power plants to create a space in the containment for the placing and cooling of the molten corium under various severe accidents. This research investigates the role of the core catcher in the VVER-1000 reactor containment system in mitigating the effects of core meltdown under various severe accidents within the context of the Ex-vessel Melt Retention (EVMR) strategy. Hence, a comparison study of three severe accidents is conducted, including Station Black-Out (SBO), SBO combined with the Large Break Loss of Coolant Accident (LB-LOCA), and SBO combined with the Small Break Loss of Coolant Accident (SB-LOCA). Numerical comparative simulations are performed for the aforementioned scenario with and without the EX-vessel core-catcher. The results showed that considering the EX-Vessel core catcher reduces the amount of hydrogen by about 18.2 percent in the case of SBO + LB-LOCA, and hydrogen production decreases by 12.4 percent in the case of SBO + SB-LOCA. Furthermore, in the presence of an EX-Vessel core-catcher, the production of gases such as CO and CO2 for the SBO accident is negligible. It was revealed that the greatest decrease in pressure and temperature of the containment is related to the SBO accident.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.37-50
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• 2020

A passive containment cooling system (PCCS) has been developed as advanced safety feature for innovative power reactor (iPOWER). Passive systems are inherently less stable than active systems and the PCCS encountered the flashing-induced instability previously identified. The objective of this study is to develop stability maps for flashing-induced instability using MARS (Multi-dimensional Analysis of Reactor Safety) code. Firstly, we conducted a series of sensitivity analysis to see the effects of time step size, nodalization, and alternative MARS user options on the onset of flashing-induced instability. The riser nodalization strongly affects the prediction of flashing in a long riser of the PCCS, while time step size and alternative user options do not. Based on the sensitivity analysis, a standard input and an analysis methodology were set up to develop the stability maps of PCCS. We found out that the calculated equilibrium quality at the exit of the riser as a stability boundary above 5 kW/㎡ was approximately 1.2%, which was in good agreement with Furuya's results. However, in case of a very low heat flux condition, the onset of instability occurred at the lower equilibrium quality. In addition, it was confirmed that inlet throttling reduces the unstable region.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2488-2498
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• 2021

The wall film-wise condensation plays an important role in the heat transfer processes of heat exchangers, refrigerators, and air conditioner. In the field of nuclear engineering, steam condensation is often utilized in safety systems to remove the core decay heat under both transient and accident conditions. In particular, passive containment cooling system (PCCS), are designed to ensure containment safety under severe accident conditions. A computational fluid dynamics (CFD) scale analysis has been conducted to calculate the heat transfer rate of the PCCS. However, despite the increase in computing power, there are challenges in the long-term transient simulation of containment using CFD scale codes. In this study, a heat structure coupling between the CFD and system analysis codes was performed to efficiently analyze PCCS. In addition, the component unstructured program for interfacial dynamics (CUPID) was improved to analyze the condensation behavior of ternary gas mixtures. Thereafter, the condensation heat transfer on the primary side was calculated using the improved CUPID and CFD code, whereas that on the secondary side was simulated using MARS. Both the coupled codes were validated against the CONAN facility database. Finally, conjugate heat transfer simulations with wall condensation in the presence of non-condensable gases were appropriately performed.