• 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.

• Journal of Korean Society for Quality Management
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• v.21 no.2
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• pp.242-253
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• 1993

The interpretation of the reaction mechanism is significant to produce the high quality welds and understand the welding processes. This investigation is important for the design of welding consumables and the selection of welding process parameters to develop the high quality welds. The objective of this study is to investigate the effect of electrochemical reactions on the transfer of alloy elements between slag and weld metal during submerged arc welding During submerged arc welding weld metal composition is shown to be controlled by two reaction mechanisms in four reaction zones. The responsible reaction mechanisms are thermochemical and electrochemical reactions. The possible reaction sites are the melted electrode tip, the detached droplet, the hot weld pool immediately below the moving electrode, and the cooling and solidifying weld pool behind the moving electrode. The possible reactions in submerged arc welding at different zones of the process is schematically shown in Figure 1.

• Nuclear Engineering and Technology
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• v.17 no.1
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• pp.1-7
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• 1985

Non-destructive gamma-ray spectrometry was carried out on the spent PWR fuel assemblies at the spent fuel pool of reactor-site. Attention was focused on the determination of burnup distribution and cooling time. For the measurement of burnup distribution, the concentration ratio of $^{134}$ Cs$^{137}$ Cs was used and the results showed these ratios varied with the positions of assemblies in the core during their irradiation. For the measurement of cooling time, $^{144}$ Ce$^{137}$ Cs was used and the results were agreed considerably well with the operator declared cooling time.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.34 no.7
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• pp.935-941
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• 2010

HANARO is an open-tank-in-pool-type Korean research reactor that generates 30MW of thermal power. It differs from power plant reactor in that the heat generated by HANARO is exhausted into the atmosphere through a secondary cooling tower, thus maintaining the core temperature constant. During every monthly inspection of the cooling tower, large vibrations that exceeded the permissible limit were observed at cooling fan gear reducer No. 4 of the cooling tower. The purpose of this study is to identify the origin of the large vibration and to repair it. FFT spectrum analysis is performed to identify the part that caused the large vibration. The results of the frequency analysis showed that the vibration frequency was 354 Hz, which is twice the natural frequency of the pinion gear. A check of the pinion gear revealed that there was a crack on the surface of the pinion gear. After the gear was replaced, the reducer operated normally.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 1993.10a
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• pp.518-523
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• 1993

A computer program has been developed for analyzing the two-dimensional unsteady conservation equations for transport phenomena in the ool region of direct rolling with semi-solid metal in order to describe the velocity and temperature, and the solidification process of the semi-solid metal. The energy equations of cooling roll is solved simultaneously with semi-solid metal in order consider heat transfer through the cooling roll. The FDM(finite difference method) and FEM(finite element method) are used in region of pool and roll, respsctively, to reduce computing time and to improve accuracy of calculation. In the present study, influence of solid fraction and casting speed are investigated in a point of view of strip formability with semi-solid metal.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.767-790
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• 2013

MAAP4 is a computer code that can simulate the response of a light water reactor power plant during severe accident sequences, including actions taken as part of accident management. The code quantitatively predicts the evolution of a severe accident starting from full power conditions given a set of system faults and initiating events through events such as core melt, reactor vessel failure, and containment failure. Furthermore, models are included in the code to represent the actions that could mitigate the accident by in-vessel cooling, external cooling of the reactor pressure vessel, or cooling the debris in containment. A key element tied to using a code like MAAP4 is an uncertainty analysis. The purpose of this paper is to present a MAAP4 based analysis to examine the sensitivity of a key parameter, in this case hydrogen production, to a set of model parameters that are related to a Level 2 PRA analysis. The Level 2 analysis examines those sequences that result in core melting and subsequent reactor pressure vessel failure and its impact on the containment. This paper identifies individual contributors and MAAP4 model parameters that statistically influence hydrogen production. Hydrogen generation was chosen because of its direct relationship to oxidation. With greater oxidation, more heat is added to the core region and relocation (core slump) should occur faster. This, in theory, would lead to shorter failure times and subsequent "hotter" debris pool on the containment floor.

• Nuclear Engineering and Technology
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• v.47 no.1
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• pp.66-73
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• 2015

Background: A numerical simulation was carried out to investigate the difference between internal and external heat-flux distributions at the reactor vessel wall under in-vessel retention through external reactor vessel cooling (IVR-ERVC). Methods: Total loss of feed water, station blackout, and large break loss of coolant accidents were selected as the severe accident scenarios, and a transient analysis using the element-birth-and-death technique was conducted to reflect the vessel erosion (vessel wall thickness change) effect. Results: It was found that the maximum heat flux at the focusing region was decreased at least 10% when considering the two-dimensional heat conduction at the reactor vessel wall. Conclusion: The results show that a higher thermal margin for the IVR-ERVC strategy can be achieved in the focusing region. In addition, sensitivity studies revealed that the heat flux and reactor vessel thickness are dominantly affected by the molten corium pool formation according to the accident scenario.

• Fire Science and Engineering
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• v.23 no.6
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• pp.57-65
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• 2009

Fire suppression tests are carried out for a train car using water mist systems. Three kinds of fire scenario applied to the real-scale train car are a surface fire representing car combustibles, a oil pool fire pretending an oil spill and a blocked fire for evaluation of space-cooling capacity. Five fixedpressure water mist systems and one self-contained water mist system with nitrogen gas are used for fire suppression experiments. Almost water mist systems can extinguish effectively train car fires, and fire-control capability of the system is seen due to the space cooling.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.3
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• pp.303-313
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• 2023

Spent fuels (SFs) are stored in a storage pool after discharge from nuclear power plants. They can be transferred to for the further processes such as dry storage sites, processing plants, or disposal sites. One of important measures of SF is the burnup. Since the radioactivity of SF is strongly dependent on its burnup, the burnup of SF should be well estimated for the safe management, storage, and final disposal. Published papers about the methodology for the burnup estimation from the known activities of important radioactive sources are somewhat rare. In this study, we analyzed the dependency of the burnup on the important radiation source activities using ORIGEN-ARP, and suggested simple correlations that relate the burnup and the important source activities directly. A burnup estimation equation is suggested for PWR fuels relating burnup with total neutron source intensity (TNSI), initial enrichment, and cooling time. And three burnup estimation equations for major gamma sources, ¹³⁷Cs, ¹³⁴Cs, and ¹⁵⁴Eu are also suggested.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.3017-3029
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• 2023

In the late in-vessel phase of a nuclear reactor severe accident, the internal heat transfer and crust evolution during the debris bed melting process have important effects on the thermal load distribution along the vessel wall, and further affect the reactor pressure vessel (RPV) failure mode and the state of melt during leakage. This study coupled the phase change model and large eddy simulation to investigate the variations of the temperature, melt liquid fraction, crust and heat flux distributions during the debris bed melting process in the hypothetical severe accident of HPR1000. The results indicated that the heat flow towards the vessel wall and upper surface were similar at the beginning stage of debris melting, but the upward heat flow increased significantly as the development of the molten pool. The maximum heat flux towards the vessel wall reached 0.4 MW/m². The thickness of lower crust decreased as the debris melting. It was much thicker at the bottom region with the azimuthal angle below 20° and decreased rapidly at the azimuthal angle around 20-50°. The maximum and minimum thicknesses were 2 and 90 mm, respectively. By contrast, the distribution of upper crust was uniform and reached stable state much earlier than the lower crust, with the thickness of about 10 mm. Moreover, the sensitivity analysis of initial condition indicated that as the decrease of time interval from reactor scram to debris bed dried-out, the maximum debris temperature and melt fraction became larger, the lower crust thickness became thinner, but the upper crust had no significant change. The sensitivity analysis of in-vessel retention (IVR) strategies indicated that the passive and active external reactor vessel cooling (ERVC) had little effect on the internal heat transfer and crust evolution. In the case not considering the internal reactor vessel cooling (IRVC), the upper crust was not obvious.