• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2859-2874
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• 2024

This paper presents the finite element deformation and failure simulation of a typical Korean high-power reactor vessel under a severe accident characterized by large break loss of coolant (LBLOCA) with in-vessel retention of molten corium through external reactor vessel cooling (IVR-ERVC) conditions. Temperature distributions calculated using Modular Accident Analysis Program Version 5 (MAAP5) as thermal boundary conditions were used, and ABAQUS thermal and structural analyses were performed. After full ablation, the temperature of the inner surface in the thinnest section remained high (920 ℃), but the stress remained relatively low (less than 6 MPa). At the outer surface, the stress was as high as 250 MPa; however, the resulting plastic strain was small owing to the low temperature of 200 ℃. Variations in stress, inelastic strain, and temperature with time in the thinnest section suggest that the plastic and creep strains are saturated owing to stress relaxation, resulting in low cumulative damage. Thus, the lower head of the vessel can maintain its structural integrity under LBLOCA with IVR-ERVC conditions. The sensitivity analysis of internal pressure indicates the occurrence of failure in the thinnest section at an internal pressure >9.6 MPa via local necking followed by failure due to high stresses.

• Structural Engineering and Mechanics
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• v.8 no.1
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• pp.103-118
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• 1999

The structural integrity of the reactor pressure vessel under pressurized thermal shock (PTS) is evaluated in this study. For given material properties and transient histories such as temperature and pressure, the stress distribution is found and stress intensity factors are obtained for a wide range of crack sizes. The stress intensity factors are compared with the fracture toughness to check if cracking is expected to occur during the transient. A round robin problem of the PTS during a small break loss of coolant transient has been analyzed as a part of the international comparative assessment study, and the evaluation results are discussed. The maximum allowable nil-ductility transition temperatures are determined for various crack sizes.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.598-603
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• 1996

When locked rotor event. occurs, instantaneously affected loop and core flow were quickly reduced, which resulted in an increase in coolant temperature and system pressure. Analysis method of this event was that constant core inlet temperature and system pressure as well as change in core flow calculated from COAST code were statically used as an input variable to HERMITE code, because of no tools to simulate NSSS behavior and 1-D core neutronics transient coincidently. With employing TASS code revised with 1-D neutronics model, this event was analyzed in point of DNBR. By doing so, analysis procedure could be simplified and unreasonable conservatism might be removed in DNBR calculation by consideration of pressure increase.

• Proceedings of the KSME Conference
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• 2000.11a
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• pp.93-98
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• 2000

Unpredictable failures can occur due to the DHC (delayed hydride cracking) or the degradation of fracture toughness by hydride embrittlement in CANDU pressure tube which can result from the absorption of hydrogen or deuterium in the high temperature coolant. To investigate the hydride embrittlement of CANDU Zr-2.5Nb pressure tube, the transverse tensile test and the fracture toughness test were performed from room temperature to $300^{\circ}C$ using three different specimens which have an AR (As Received), 100, and 200 ppm hydrogen. As the amount of absorbed hydrogen was increased, the transverse yield strength and the ultimate tensile strength were also increased. In addition, as the test temperature became higher they were decreased linearly. While, at room temperature, the hydrogenbsorbed specimens represented the embrittlement which resulted in sudden decreasing of fracture toughness, the fracture characteristics became ductile such as AR specimen at high temperatures. Through the observation of fracture surface using SEM, it was found that the stress state of mixed mode could be related to the fissure which was believed to decrease the global fracture toughness.

• Journal of the Korean Society of Propulsion Engineers
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• v.22 no.6
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• pp.84-93
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• 2018

Three-dimensional (3D) numerical simulations have been carried out to study how coolant injection conditions influence the cooling efficiency and projectile ejection performance in a mixture-gas ejection system (or gas-steam launch system). The 3D single-phase computational model was verified using a 1D model constructed with reference to the previous research and then a two-phase flow computation simulating coolant injection on to hot gas was performed using a DPM (Discrete Phase Model). As a result of varying the coolant flow rate and number of injection holes, cooling efficiency was improved when the number of injection holes were increased. In addition, the change of the coalescence frequency and spatial distribution of coolant droplets caused by the injection condition variation resulted in a change of the droplet diameter, affecting the evaporation rate of coolant. The evaporation was found to be a critical factor in the design optimization of the ejection system by suppressing the pressure drop while the temperature decreases inside the breech.

• Transactions of the Korean Society of Automotive Engineers
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• v.16 no.1
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• pp.37-44
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• 2008

The global warming potential (GWP) of $CO_2$ refrigerant is 1/1300 times lower than that of R134a. Furthermore, the size and weight of the automotive heat pump system can decrease because $CO_2$ operates at high pressure with significantly higher discharge temperature and larger temperature change. The presented $CO_2$ heat pump system was designed for both cooling and heating in fuel cell vehicles. In this study, the performance characteristics of the heat pump system were analyzed for heating, and results for performance were provided for operating conditions when using recovered heat from the stack coolant. The performance of the heat pump system with heater core was compared with that of the conventional heating system with heater core and that of the heat pump system without heater core, and thus the heat pump system with heater core showed the best performance among the selected heating systems. On the other hand, the heating performance of two different types of coolant/air heat pump systems with heater core was compared each other at various coolant inlet temperatures. Furthermore, to use exhausted thermal energy through the radiator, experiments were carried out by changing the arrangement of a radiator and an outdoor evaporator, and quantified the heating effectiveness.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.20 no.1
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• pp.75-83
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• 2024

In this study, high temperature design and integrity evaluation methodology have been developed for Type 316L stainless steel air-to-sodium heat exchanger which uses 700℃ sodium as coolant. Currently the only design rules that take creep effects into consideration explicitly for the 316L stainless steel subjected to high temperature in the creep range are French RCC-MRx, where elevated temperature designs are possible around 550℃. Absent design coefficients at high temperature were determined based on the material properties newly determined in previous studies, and high-temperature design evaluation methodologies were developed based on 3D finite element analyses on the 700℃ class heat exchanger. In addition, works were conducted on the web-based design evaluation program of HITEP_RCC-MRx including incorporation of material properties and design coefficients up to 700℃. Methodologies on high temperature design evaluations on Type 316L stainless steel high-temperature heat exchanger were suggested.

• 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.48 no.4
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• pp.893-905
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• 2016

Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2002.05a
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• pp.24-27
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• 2002

The environmental problems by using coolant demanded the new cooling methods. As one of them, the studies on the finding with compressed cold air and oil mist have been done. The cooling method using compressed cold air was effective through going down the temperature of compressed air supplied below $-25^{\circ}$ and increasing the amount of compressed cold air, but had not enough cooling effect due to the low performance of lubrication. Therefore, the cooling methods using oil mist newly were suggested. This method can satisfy both cooling effect and lubrication with only small amount of coolant, also have the benefit in the point of decreasing the environmental pollution. This paper focused on analyzing the grinding characteristics of the cooling method using oil mist. The grinding test according to compressed cold air, oil mist spray pressure and oil mist supply direction were done.