• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2499-2508
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• 2021

Passive containment cooling systems (PCCSs) have been actively studied to improve the inherent safety of nuclear power plants. Hered, we present two concepts, open-loop PCCS (OL-PCCS) and closed-loop PCCS (CL-PCCS), applicable to the PWR with a concrete-type containment. We analyzed the heat-removal performance and flow instability of these PCCS concepts using the GOTHIC code. In both cases, PCCS performance improved when a passive containment cooling heat exchanger (PCCX) was installed in the lower part of the containment building. The OL-PCCS was found to be superior in terms of heat-removal performance. However, in terms of flow instability, the OL-PCCS was more vulnerable than the CL-PCCS. In particular, the possibility of flow instability was higher when the PCCX was installed in the upper part of the containment. Therefore, the installation location of the OL-PCCS should be restricted to minimize flow instability. Conversely, a CL-PCCS can be installed without any positional restriction by adjusting the initial system pressure within the loop, which eliminates flow instability. These results could be used as base data for the thermo-hydraulic evaluation of PCCS in PWR with a large dry concrete-type containment.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.358-363
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• 1995

A study on passive cooling systems for concrete containment of advanced pressurized water reactors has been performed. The proposed passive containment cooling system (PCCS) consist of (1) condenser units located inside containment, (2) a steam condensing pool outside containment at higher elevation, and (3) downcommer/riser piping systems which provide coolant flow paths. During an accident causing high containment pressure and temperature, the steam/air mixture in containment is condensed on the outer surface of condenser tubes transferring the heat to coolant flowing inside tubes. The coolant transfers the heat to the steam condensing pool via natural circulation due to density difference. This PCCS has the following characteristic: (1) applicable to concrete containment system, (2) no limitation in plant capacity expansion, (3) efficient steam condensing mechanism (dropwise or film condensation at the surface of condenser tube), and (4) utilization of a fully passive mechanism. A preliminary conceptual design work has been done based on steady-state assumptions to determine important design parameter including the elevation of components and required heat transfer area of the condenser tube. Assuming a decay power level of 2%, the required heat transfer area for 1,000MWe plant is assessed to be about 2,000 ㎡ (equivalent to 1,600 of 10 m-long, 4-cm-OD tubes) with the relative elevation difference of 38 m between the condenser and steam condensing pool and the riser diameter of 0.62 m.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1134-1145
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• 2021

A design parameter study is presented for the closed-loop type passive containment cooling system (PCCS) which is equipped with two heat exchangers: one installed at the inside of the containment and the other submerged in the water pool at the outside of the containment. A GOTHIC code model for PCCS performance analyses was set up and the design parameters such as the heat exchanger sizes, locations, and water pool tank volumes were analyzed to investigate the feasibility of installing this type of PCCS in PWRs like OPR-1000 being operated in Korea. We identified the size of the circulation loop and heat exchangers as major design parameters affecting the performance of PCCS. The analyses showed that the heat exchangers in the inside of the containment would be more influential on the heat removal capability of PCCS than that installed in the water pool at the outside of the containment. Hence, it was recommended to down-size the heat exchangers in the water pool to optimize PCCS without compromising its performance. Based on the parametric study, it was demonstrated that a closed-loop type PCCS could be designed sufficiently compact for installation in the available space within the containment of PWRs like OPR-1000.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.268-273
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• 2016

An innovative safety concept for a light water reactor has been developed at the Korea Atomic Energy Research Institute. It is a unique concept that adopts both a fast heat transfer mechanism for a small containment and a changing mechanism of the cooling geometry to take advantage of the potential, thermal, and dynamic energies of the cold water in the containment. It can bring about rapid cooling of the containment and long-term cooling of the decay heat. By virtue of this innovative concept, nuclear fuel damage events can be prevented. The ultimate heat transfer mechanism contributes to minimization of the heat exchanger size and containment volume. A small containment can ensure the underground construction, which can use river or seawater as an ultimate heat sink. The changing mechanism of the cooling geometry simplifies several safety systems and unifies diverse functions. Simplicity of the present safety system does not require any operator actions during events or accidents. Therefore, the unique safety concept of PX can realize both economic competitiveness and inherent safety.

• 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.49 no.3
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• pp.459-465
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• 2017

The goal of this study is to evaluate the effective thermal conductivity and diffusivity of containment walls as heat sinks or passive cooling systems during nuclear power plant (NPP) accidents. Containment walls consist of steel reinforced concrete, steel liners, and tendons, and provide the main thermal resistance of the heat sinks, which varies with the volume fraction and geometric alignment of the rebar and tendons, as well as the temperature and chemical composition. The target geometry for the containment walls of this work is the standard Korean NPP OPR1000. Sample tests and numerical simulations are conducted to verify the correlations for models with different densities of concrete, volume fractions, and alignments of steel. Estimation of the effective thermal conductivity and diffusivity of the containment wall models is proposed. The Maxwell model and modified Rayleigh volume fraction model employed in the present work predict the experiment and finite volume method (FVM) results well. The effective thermal conductivity and diffusivity of the containment walls are summarized as functions of density, temperature, and the volume fraction of steel for the analysis of the NPP accidents.

• 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.

• Journal of Energy Engineering
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• v.24 no.1
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• pp.42-50
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• 2015

To evaluate the heat removal capability of a condenser tube in the PCCS of an advanced nuclear power plant, a steam condensation experiment in the presence of noncondensable gas on a vertical tube is performed. The average heat transfer coefficient is measured on a vertical tube of 40 mm in O.D. and 1.0 m in length. The experiments covers the pressures of 2-4 bar, and the mass fraction of air ranges from 0.1 up to 0.7. From the experimental results, the effects of the total pressure and the concentration of air on the condensation heat transfer coefficient are investigated. The measured data are compared with the predictions by Uchida's and Tagami's correlations, and it is revealed that these models underestimate the condensation heat transfer coefficient of the steam-air mixture.

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.605-612
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• 2013

Removal of decay heat from an operating reactor during a prolonged station blackout condition is a big concern for reactor designers, especially after the recent Fukushima accident. In the case of a prolonged station blackout condition, heat removal is possible only by passive means since no pumps or active systems are available. Keeping this in mind, the AHWR has been designed with many passive safety features. One of them is a passive means of removing decay heat with the help of Isolation Condensers (ICs) which are submerged in a big water pool called the Gravity Driven Water Pool (GDWP). The ICs have many tubes in which the steam, generated by the reactor core due to the decay heat, flows and condenses by rejecting the heat into the water pool. After condensation, the condensate falls back into the steam drum of the reactor. The GDWP tank holds a large amount of water, about 8000 $m^3$, which is located at a higher elevation than the steam drum of the reactor in order to promote natural circulation. Due to the recent Fukushima type accidents, it has been a concern to understand and evaluate the capability of the ICs to remove decay heat for a prolonged period without escalating fuel sheath temperature. In view of this, an analysis has been performed for decay heat removal characteristics over several days of an AHWR by ICs. The computer code RELAP5/MOD3.2 was used for this purpose. Results indicate that the ICs can remove the decay heat for more than 10 days without causing any bulk boiling in the GDWP. After that, decay heat can be removed for more than 40 days by boiling off the pool inventory. The pressure inside the containment does not exceed the design pressure even after 10 days by condensation of steam generated from the GDWP on the walls of containment and on the Passive Containment Cooling System (PCCS) tubes. If venting is carried out after this period, the decay heat can be removed for more than 50 days without exceeding the design limits.