• Journal of the Korean Society of Safety
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• v.16 no.4
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• pp.29-38
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• 2001

The purpose of this paper is to evaluate the structural integrity of a reactor pressure vessel subjected to the pressurized thermal shock(PTS) during the transient events, such as main steam line break(MSLB) and small break loss of coolant accident(SBLOCA). For postulated surface or subsurface cracks, variation curves of stress intensity factor are obtained by using the three different methods, including ASME section XI code anlysis, the finite element alternating method and the finite element method. From the stress intensity factor curves, the maximum allowable nil-ductility transition temperatures(RT/NDT/) are determined by the tangent criterion and the maximum criterion for various crack configurations and two initial transient events. As a result of the analysis, it is noted that axial cracks have smaller maximum allowable RT$_{NDT}$ values than same-sized circumferential cracks for both the transient events in the case of the tangent criterion. Axial cracks have smaller RT$_{NDT}$ values than same-sized circumferential cracks for MSLB and circumferential cracks have smaller values than axial cracks for SBLOCA in the case of the maximum criterion.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.397-398
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• 2004

It is important to estimate the radiation activity of the nuclear fuel which is a source term of the loss of coolant accident. The purpose of this study is to identify the most important parameters of the source term calculation based on three fuel types: typical natural uranium CANDU fuel, slightly enriched uranium and DUPIC fuel. The characteristics of the radiation source term were analyzed through sensitivity calculations of the linear power, fuel turnup, and the power shape.(omitted)

• Nuclear Engineering and Technology
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• v.14 no.4
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• pp.204-218
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• 1982

The purpose of this report is to describe and discuss the current state of the knowledge about the Leidenfrost phenomenon which is a heavily studied subject in the field of boiling heat transfer. The strong interest is due to reactor safety considerations since it is desirable to obtain a better understanding of the physical mechanisms involved in the rewetting of high temperature surface after a loss of coolant accident. Brief survey of the theoretical and experimental results from available open literatures indicates that considerable discrepancy exists in the prediction of the Leidenfrost temperature at the elevated pressures and more investigations are needed in this area.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.635-640
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• 1998

Thermal fluid mixing behavior during a postulated medium-size hot leg break loss of coolant accident is analyzed for the international comparative assessment study on pressurized thermal shock (PTS-ICAS) proposed by OECD-NEA. The applicability of RELAP5 code to analyze thermal fluid mixing behavior is evaluated through a simple modeling relevant to the problem constraints. Based on the calculation result, the onset of Thermal stratification is investigated using Theofanous's empirical correlation. Sensitivity calculations using a fine node model and crossflow model are also performed to evaluate the modeling capability on multi-dimensional characteristics related to thermal fluid mixing.

• Nuclear Engineering and Technology
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• v.32 no.5
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• pp.521-529
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• 2000

A modified interfacial heat transfer correlation between a dispersed water droplet and ambient superheated steam is proposed and compared with available experimental data and other correlations. Modified one overcomes the inherent deficiencies of Lee and Ryley's interfacial heat transfer correlation that ignored the effects of steam superheating which can not be neglected especially in the reflood situation of a loss-of-coolant accident. Modified one is represented by (equation omitted) In the present correlation the effect of possible subcooling of a water droplet is not taken into consideration. Comparison of the above correlation with currently available measurement data for a water droplet in high temperature gas flow shows that the proposed one correlates well with the measurement data where the degree of superheating is negligible and considerable.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.785-792
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• 2024

This article provides an overview of the design methodology used to develop a conceptual set of reactivity control mechanism of a micro reactor based on the U-Battery. The U-Battery is based on remote deployment and therefore it is favourable to provide a long fuel lifecycle. This is achieved by implementing a high fissile loading content, which proves challenging when considering reactivity control methods. This article follows the design methodology used to overcome these issues, with an emphasis on a new concept of a moveable moderator which utilises the size of the U-Battery as a small reduction in moderation provides a significant reduction in reactivity. The latest work on this project sees the moveable moderator investigated during a depressurised loss of forced coolant accident, where a reduction of moderator volume increases the maximum fuel temperature experienced. The overall conclusion is that the maximum fuel temperature is not significantly increased (4 K) due to the central reflector region relatively lower volumetric heat capacity compared to that of whole core. However, a small temperature increase is observed immediately after the transient due to the central reflector removal because it reaches energy equilibrium with the fuel region faster.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.838-843
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• 1998

Risk importance measures are widely wed to rank risk contributors in risk-based applications. Typically, Fussell-Vesely (F-V) importance and risk achievement worth (RAW) are used in the component importance raking for the reliability centered maintenance (RCM) analysis of safety system in nuclear power plants (NPPs). This study was performed as part of feasibility study on RCM for domestic NPPs, which is focused on the component importance ranking approach the maintenance recommendation. The approach of modulizing faulting tree basic events was applied in the simplification process of the PSA model and the validity of the approach was evaluated As a result of the case study, this paper included the importance and the maintenance recommendations for the safety-related equipments associated with safety injection and containment spray in large loss of coolant accident sequences.

• 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.

• Journal of the Korean Society for Heat Treatment
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• v.17 no.4
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• pp.216-222
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• 2004

Study was focused on the quenching embrittlement property of Zircaloy-4 cladding simulated Loss Of Coolant Accident (LOCA) environment in terms of high temperature oxidation and phase transformation. Property in LOCA condition of advanced cladding that contained Nb element was also investigated. Claddings were oxidized at given temperature and given time followed by water quenching. The results showed that ${\beta}$ phase which formed at quenching stage has an influence on cladding property. In case of advanced cladding, Nb retards cladding oxidation, thus enhances quenching resistance.

• Nuclear Engineering and Technology
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• v.15 no.2
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• pp.98-109
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• 1983

A two-step alternating direction explicit method is developed to solve the space-dependent reactor kinetics equations in two space dimensions. As a special case in the general class of alternating direction implicit methods, this method is analysed for accuracy and stability. To test the validity of this method it is compared with the implicit-difference method used in the TWIGL program. It is shown that the two methods are closely related. The time dependent neutron fluxes of the pressurized water reactor (PWR), during control rod insertion, and, of the CANDU-PHW reactor, in case of postulated loss of coolant accident, are obtained from the numerical calculation results.