• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2477-2487
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• 2021

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

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.25 no.2
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• pp.79-84
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• 2013

In this study, a cold aisle containment system was proposed among various strategies to reduce the energy waste by recirculation air from the hot aisle. To verify the effectiveness of the cold aisle containment system, a test bed which is similar to an actually existing server room was set up in the Internet Data Center(IDC) building. Comparative experiments, conventional open type cooling system and cold aisle containment system were carried out under actual conditions. The result revealed that the range of inlet temperature of the server system was $20{\sim}25^{\circ}C$ in an existing cooling system and the range of inlet temperature dropped below $20^{\circ}C$ by the cold aisle containment system. After all, cold aisle containment system was proved to be the solution for energy saving cooling system.

• 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.

• Proceedings of the KIEE Conference
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• 2005.10b
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• pp.490-492
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• 2005

Unattended continuous containment/surveillance systems for safeguards of nuclear facility result in large amounts of image and radiation data, which require much time and effort to inspect. Therefore, it is necessary to develop system that automatically pinpoints and diagnoses the anomalies from data. In this regards, this paper presents the nuclear material containment/surveillance system that integrates visual image and radiation data.

• Journal of ILASS-Korea
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• v.4 no.3
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• pp.42-52
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• 1999

An effectiveness of thermal utilization of a dousing system in the 600 MW PHWR Nuclear Power Plant has been evaluated. The behavior and conditions of water droplet sprayed in a postulated accident conditions in containment configuration has been calculated. In this calculation, two pressure conditions with the consideration of obstruction area and containment wall effect has been established : one being the minimum containment pressure of 7 kPa(g) encountered for dousing shut off and the other being the containment design pressure 124 kPa(g). The results revealed that the effectiveness of the thermal utilization ranges from 93% to 97%. In the analysis on two cases without/with side wall effect in the containment building, the thermal utilization decreases with obstruction area from 89% to 85%, which satisfies the design criteria set for the containment pressure against the accident condition.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3196-3206
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• 2021

For the Korean design of the PCCS (passive containment cooling system) in an innovative PWR, the overall thermal resistance around a condenser tube is dominated by the heat transfer coefficient of steam condensation on the exterior surface. It has been reported, however, that the calculated heat transfer coefficients by thermal-hydraulic system codes were much lower than measured data in separate effect tests. In this study, a new empirical model of steam condensation in the presence of a noncondensable gas was implemented into the MARS-KS 1.4 code to replace the conventional Colburn-Hougen model. The selected correlation had been developed from condensation test data obtained at the JERICHO (JNU Experimental Rig for Investigation of Condensation Heat transfer On tube) facility, and considered the effect of the Grashof number for naturally circulating gas mixture and the curvature of the condenser tube. The modified MARS-KS code was applied to simulate the transient response of the containment equipped with the PCCS to the large-break loss-of-coolant accident. The heat removal performances of the PCCS and corresponding evolution of the containment pressure were compared to those calculated via the original model. Various thermal-hydraulic parameters associated with the natural circulation operation through the heat transport circuit were also investigated.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1513-1525
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• 2024

An improved mitigation system for thermally-induced steam generator tube rupture accidents was introduced to prevent direct environmental release of fission products bypassing the containment in the OPR1000. This involves injecting bypassed steam into the containment, cooling, and decontaminating it using a water coolant tank. To evaluate its performance, a severe accident analysis was performed using the MELCOR 2.2 code for OPR1000. Simulation results show that the proposed system sufficiently prevented the release of radioactive nuclides (RNs) into the environment via containment injection. The pool scrubbing system effectively decontaminated the injected RN and consequently reduced the aerosol mass in the containment atmosphere. However, the decay heat of the collected RNs causes re-vaporization. To restrict the re-vaporization, an external water source was considered, where the decontamination performance was significantly improved, and the RNs were effectively isolated. However, due to the continuous evaporation of the feed water caused by decay heat, a substantial amount of steam is released into the containment. Despite the slight pressurization inside the containment by the injected and evaporated steam, the steam decreased the hydrogen mole fraction, thereby reducing the possibility of ignition.

• Proceedings of the KIEE Conference
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• 2004.11c
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• pp.635-637
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• 2004

The containment of the nuclear power plant is the last barrier of radiation release when the reactor coolant pipe rupture is occurred. Each plant has to be tested every 5 years whether the containment leak rate meets its technical specifications. We have developed the leak rate test system and in this paper, we describe the results of the uncertainty analysis on the measurement channels and its propagation to the calculation results.

• 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.

• 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.