• Journal of Mechanical Science and Technology
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• v.17 no.9
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• pp.1380-1387
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• 2003

When a cold HPSI (High pressure Safety Injection) fluid associated with an overcooling transient, such as SGTR (Steam Generator Tube Rupture), MSLB (Main Steam Line Break) etc., enters the cold legs of a stagnated primary coolant loop, thermal stratification phenomena will arise due to incomplete mixing. If the stratified flow enters the downcomer of the reactor pressure vessel, severe thermal stresses are created in a radiation embrittled vessel wall by local overcooling. As general thermal-hydraulic system analysis codes cannot properly predict the thermal stratification phenomena, RG 1.154 requires that a detailed thermal-mixing analysis of PTS (pressurized Thermal Shock) evaluation be performed. Also. previous PTS studies have assumed that the thermal stratification phenomena generated in the stagnated loop side of a partially stagnated primary coolant loop are neutralized in the vessel downcomer by the strong flow from the unstagnated loop. On the basis of these reasons, this paper focuses on the development of a 3-dimensional thermal-mixing analysis model using PHOENICS code which can be applied to both partial and total loop stagnated cases. In addition, this paper verifies the fact that, for partial loop stagnated cases, the cold plume generated in the vessel downcomer due to the thermal stratification phenomena of the stagnated loop is almost neutralized by the strong flow of the unstagnated loop but is not fully eliminated.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.24 no.8 s.179
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• pp.2083-2090
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• 2000

In order to manage nuclear power plants safely and cost effectively, it is necessary to develop integrity evaluation methodologies for the main components. Recently, the integrity evaluation techniques were broadly studied regarding the license renewal of nuclear power plants which were approaching their design lives. Since the integrity evaluation process requires special knowledges and complicated calculation procedures, it has been allowed only to experts in the specified area. In this paper, an integrity evaluation system for reactor pressure vessel was developed. RVIES(Reactor Vessel Integrity Evaluation System) provides four specific integrity evaluation procedures covering PTS(Pressurized Thermal Shock) analysis, P-T(Pressure-Temperature) limit curve generation, USE(Upper Shelf Energy) analysis and Fatigue analysis. Each module was verified by comparing with published results.

• Journal of Hydrogen and New Energy
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• v.31 no.2
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• pp.184-193
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• 2020

One of the technical methods to increase the volumetric energy density of hydrogen is to pressurize the gaseous hydrogen and then contain it in a rigid vessel. Especially for automotive systems, the compressed hydrogen storage can be found in cars as well as at refueling stations. During the charging the pressurized hydrogen into a vessel, the temperature increases with the amount of stored hydrogen in the vessel. The temperature of the vessel should be controlled to be less than a limitation for ensure stability of material. Therefore, the accurate estimation of temperature is of significance for safely storing the hydrogen. In this work, three well-known cubic equations of state (EOSs) were adopted to examine the accuracy in regenerating thermodynamic properties of hydrogen within the temperature and pressure ranges for the compressed hydrogen storage. The formulations representing molar volume, internal energy, enthalpy, and entropy were derived for Redlich-Kwong (RK), Soave-Redlioch-Kwong (SRK), and Peng-Robinson (PR) EOSs. The calculated results using the EOSs were compared with literature data given by NIST. It was revealed that the accuracies of RK and SRK EOSs were satisfactorily compatible and better than the results by PR EOS.

• Nuclear Engineering and Technology
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• v.56 no.8
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• pp.3043-3066
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• 2024

In various small modular reactor (SMR) designs currently under development, the conventional concrete containment building has been replaced by a metal containment vessel (MCV). In these systems, the gap between the MCV and the reactor pressure vessel is filled with gas or vacuumed weakly, effectively suppressing conduction and convection heat transfer. However, thermal radiation remains the major mode of heat transfer during normal operation. The objective of this study was to investigate the heat-transfer mechanisms in integral pressurized water reactor (IPWR)-type SMRs under various gas-filled conditions using computational fluid dynamics. The use of thermal radiation shielding (TRS) with a much lower emissivity material than the MCV surface was also evaluated. The results showed that thermal radiation was always the dominant contributor to heat loss (48-97%), while the conjugated effects of the gas candidates on natural convection and thermal radiation varied depending on their thermal and radiative properties, including absorption coefficient. The TRS showed an excellent insulation performance, with a reduction in the total heat loss of 56-70% under the relatively low temperatures of the IPWR system, except for carbon dioxide (13%). Consequently, TRS can be utilized to enhance the thermal efficiency of SMR designs by suppressing the heat loss through the MCV.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.59 no.10
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• pp.1889-1893
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• 2010

Differential pressure type level measuring systems have been using widely for industrial applications like drum level measurements in power plants. Because of difficulties in specific gravity compensation for vapor and liquid inside the vessel and the sensing lines, this type of measuring systems reveal significant measuring error. In this paper, the major reason causing errors on the differential pressure type level measurement is analyzed and a method of more accurate calculation for specific gravity compensation is introduced.

• 한국연소학회:학술대회논문집
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• 2014.11a
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• pp.215-218
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• 2014

High combustion efficiency of hydrogen could make it an ideal source of green energy in the future. At this time, high pressure vessel is the most reasonable method of storing hydrogen. However, such a high pressurized vessel could pose a critical threat if ruptured. For this reason, it is important to understand the mechanism of hydrogen's self-ignition when a high-pressure hydrogen released into air. This paper presents several visualization images as experimental results using high-speed camera. From the visualization images, the ignition is initiated near rupture disk immediately after failure of disk. And the initial ignition and flame is stronger as a rupture pressure increases. However, this ignition region do not affect the general self-ignition mechanism when a high-pressure hydrogen is released into air through tue after failure of disk.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.283-288
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• 1996

Direct safety injection into the reactor vessel downcomer annulus(DVI) is a fundamental feature of the KNGR(Korean Next Generation Reactor) four-train safety injection system. The numerical analysis of thermal mixing of ECC(Emergency Core Cooling) water through DVI with the water in the RVDC(Reactor Vessel Downcomer) annulus has been performed, in order to study the impact of nozzle location on the pressurized thermal shock and safety analysis. The results of this study show that the thermal mixing due to the natural circulation induced by the limiting accident conditions is sufficient to prevent temperature in the RVDC from dropping to the level of concern for PTS. When the DVI nozzle is located right above the cold leg, the temperature distribution at the outlet of flow field is most uniform. The tool used for numerical analysis is CFDS-FLOW3D.

• Nuclear Engineering and Technology
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• v.41 no.6
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• pp.775-784
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• 2009

This paper summarizes the tests performed in the ATLAS facility during its first two years of operation (2007${\sim}$2008). Two categories of tests have been performed successfully: (a) the reflood phase of the large-break loss-of-coolant accidents in a cold leg, and (b) the breaks in one of four direct vessel injection lines. Those tests contributed to understanding the unique thermal-hydraulic behavior, resolving the safety-related concerns and providing an evaluation of the safety analysis codes and methodology for the advanced pressurized water reactor, APR1400. Several important and interesting phenomena have been observed during the tests. In most cases, the ATLAS shows reasonable accident characteristics and conservative results compared with those predicted by one-dimensional safety analysis codes. A wide variety of small-break LOCA tests will be performed in 2009.

• Nuclear Engineering and Technology
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• v.43 no.2
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• pp.167-174
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• 2011

A pressurizer in a small integral type pressurized water reactor is located inside the upper region of the reactor vessel, and uses a space between the upper head of the reactor vessel and the upper region of the upper guide structure which is partially filled with fluid depending on the operating power. This new design requires a comprehensive investigation of vibration characteristics. This study investigates the modal characteristics of a pressurizer which uses a simplified cylindrical shell model, focusing on how having fluid in the shell affects vibration and response characteristics. In addition, an analysis of sloshing is performed and the response characteristics are addressed.

• Nuclear Engineering and Technology
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• v.23 no.2
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• pp.183-199
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• 1991

Transient response of temperature distributions in a Pressurized Water Reactor (PWR) pressurizer vessel wall for the Inadvertent Auxiliary Spray transient has been analyzed with conservatism accounted for the resulting thermal stresses in the regions of the vessel wall which are wetted by the spray water droplets. In order to determine the forced convective heat transfer coefficient at the inner boundary surface of vessel wall where the droplets impinge on and flow down, the transient temperatures of spray droplets when they reach the inner surface of the vessel wall after travelling from the spray nozzle through the pressurizer interior space occupied with the saturated steam-noncondensable hydrogen gas mixture have been predicted. The transient temperature distributions in the vessel wall have been obtained by using the finite element method, and the typical results have been provided. It has been shown that the results of thermal analysis are consistent with representation of the input transient and have plausible physical meaning.