• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.811-820
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• 2013

SiC has been under investigation as a potential cladding for LWR fuel, due to its high melting point and drastically reduced chemical reactivity with liquid water, and steam at high temperatures. As SiC is a brittle material its behavior during the reflood phase of a Loss of Coolant Accident (LOCA) is another important aspect of SiC that must be examined as part of the feasibility assessment for its application to LWR fuel rods. In this study, an experimental assessment of thermal shock performance of a monolithic alpha phase SiC tube was conducted by quenching the material from high temperature (up to $1200^{\circ}C$) into room temperature water. Post-quenching assessment was carried out by a Scanning Electron Microscopy (SEM) image analysis to characterize fractures in the material. This paper assesses the effects of pre-existing pores on SiC cladding brittle fracture and crack development/propagation during the reflood phase. Proper extension of these guidelines to an SiC/SiC ceramic matrix composite (CMC) cladding design is discussed.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.185-191
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• 2020

Plant-specific analyses of an advanced reactor have been performed to assure the structural integrity of the reactor pressure vessel during transient conditions, which are expected to initiate pressurized thermal shock (PTS) events. The vessel failure probabilities from the probabilistic fracture mechanics analyses are combined with the transient frequencies to generate the through-wall cracking frequencies, which are compared to the acceptance criterion. Several sensitivity analyses are performed, focusing on the orientations and sizes of cracks, the copper content, and a flaw distribution model. The results show that the integrity of the reactor vessel is expected to be maintained for long-term operation beyond the design lifetime from the PTS perspective using the design data of the advanced reactor. Moreover, a fluence level exceeding 9×10¹⁹ n/㎠ is found to be acceptable, generating a sufficient margin beyond the design lifetime.

• 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.54 no.1
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• pp.72-83
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• 2022

This paper presents a numerical investigation of two-phase natural circulation flows established when external reactor vessel cooling is applied to a severe accident of the APR1400 reactor for the in-vessel retention of the core melt. The coolability limit due to external reactor vessel cooling is associated with the natural circulation flow rate around the lower head of the reactor vessel. For an elaborate prediction of the natural circulation flow rate using a thermal-hydraulic system code, MARS-KS1.5, a three-dimensional computational fluid dynamics (CFD) simulation is conducted to estimate the flow rate and pressure distribution of a liquid-state coolant at the brink of significant void generation. The CFD calculation results are used to determine the loss coefficient at major flow junctions, where substantial pressure losses are expected, in the nodalization scheme of the MARS-KS code such that the single-phase flow rate is the same as that predicted via CFD simulations. Subsequently, the MARS-KS analysis is performed for the two-phase natural circulation regime, and the transient behavior of the main thermal-hydraulic variables is investigated.

• The KSFM Journal of Fluid Machinery
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• v.4 no.3 s.12
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• pp.53-61
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• 2001

Vortex type Fluidic Device(FD) which is installed at the bottom of Safety Injection Tank(SIT) controls the discharge flow rate from the tank. In case of loss of coolant accident the injection water flows into primary system in two steps; initial high flow rate for certain period of time and subsequent low flow rate. By two-step control of the discharge flow rate, FD can ensure the effective use of water in the tank. A small-scale FD has been tested to obtain a required flow characteristics maintaining full pressure and height of prototype, which are the major contributing parameters. Through the testing of many different arrangements of internal geometry of FD, most appropriate one was selected and its performance data was obtained. As characteristics of FD, time dependent Euler number, flow rate and pressure are presented and discussed. Also a method to predict the full size FD is presented.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.24 no.3
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• pp.338-345
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• 2000

A model testing has been performed to investigate the flow characteristics of a vortex chamber, which plays a role of a flow switch and passively controls the discharge flow rate. This method of passive flow control is a matter of concern in the design of advanced nuclear reactor systems as an alternative to the active flow control to provide emergency water to the reactor core in case of postulated accidents like LOCA (Loss-Of-Coolant Accident). By changing the inflow direction in the vortex chamber and varying the flow resistance inside the chamber, the vortex chamber can control passively the injection flowrate. Fundamental characteristics such as discharge flow rate and pressure drop of the vortex chamber are measured, and its parametric effects on the performance of the vortex chamber are also systematically investigated.

• Nuclear Engineering and Technology
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• v.49 no.2
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• pp.349-359
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• 2017

During the past 25 years, in the context of probabilistic safety assessment, efforts have been directed towards establishment of comprehensive pipe failure event databases as a foundation for exploratory research to better understand how to effectively organize a piping reliability analysis task. The focused pipe failure database development efforts have progressed well with the development of piping reliability analysis frameworks that utilize the full body of service experience data, fracture mechanics analysis insights, expert elicitation results that are rolled into an integrated and risk-informed approach to the estimation of piping reliability parameters with full recognition of the embedded uncertainties. The discussion in this paper builds on a major collection of operating experience data (more than 11,000 pipe failure records) and the associated lessons learned from data analysis and data applications spanning three decades. The piping reliability analysis lessons learned have been obtained from the derivation of pipe leak and rupture frequencies for corrosion resistant piping in a raw water environment, loss-of-coolant-accident frequencies given degradation mitigation, high-energy pipe break analysis, moderate-energy pipe break analysis, and numerous plant-specific applications of a statistical piping reliability model framework. Conclusions are presented regarding the feasibility of determining and incorporating aging effects into probabilistic safety assessment models.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.821-827
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• 2001

As one of the advanced design features of the Korea next generation reactor, direct vessel injection (DVI) system is being considered instead of conventional cold leg injection (CLl) system. It is known that the DVI system greatly enhances the reliability of the emergency core cooling (ECC) system. However, there is still a dispute on its performance in terms of water delivery to the reactor core during the reflood period of a large-break loss-of-coolant accident (LOCA). Thus, experimental validation is under progress. In this paper, a new scaling method, using time and velocity reduced linear scaling law, is suggested for the design of a scaled-down experimental facility to investigate the direct ECC bypass phenomena in PWR downcomer.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.62-69
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• 2013

A probabilistic assessment code, PRO-LOCA ver. 3.7 which was developed in an international co-operative research program, PARTRIDGE was evaluated by conducting sensitivity analysis. The effect of some variables such as simulation methods (adaptive sampling, iteration numbers, weld residual stress model), crack features(Poisson's arrival rate, maximum numbers of cracks, initial flaw size, fabrication flaws), operating and loading conditions(temperature, primary bending stress, earthquake strength and frequency), and inspection model(inspection intervals, detectable leak rate) on the failure probabilities of a surge line nozzle was investigated. The results of sensitivity analysis shows the remaining problems of the PRO-LOCA code such as the instability of adaptive sampling and unexpected trend of failure probabilities at an early stage.

• Nuclear Engineering and Technology
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• v.33 no.3
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• pp.315-326
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• 2001

Direct ECC bypass phenomena that occur in a reactor vessel downcomer with a Direct Vessel Injection (DVI) system during the reflood phase of a Large Break Loss-of-Coolant Accident (LBLOCA) are experimentally investigated using a transparent l/7.5 scaled down test facility of the Upper Plenum Test Facility (UPTF). A series of separate effect tests are peformed in order to investigate the mechanisms of direct ECC bypass and to find out its scaling parameters. Various flow regimes and phasic distribution in downcomer are identified and mapped, and the fraction of direct ECC bypass is measured under a wide range of air and water injection conditions. From the counterpart test of the UPTF Test 21-D, the dimensionless gas velocity ( $j^{*}$$_{g,eff}$) is derived experimentally, which is believed to be a major scaling parameter for the fraction of direct ECC bypass. And it is found out that the direct ECC bypass is greatly affected by the spreading width of ECC water film and the geometric configuration of the downcomer.r.