• Nuclear Engineering and Technology
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• v.30 no.5
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• pp.470-484
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• 1998

An analysis program for the evaluation of pressure vessel integrity under pressurized thermal shock (PTS) is developed. For given material properties and transient history such as temperature and pressure, the stress distribution is calculated and then 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. Using this program a round robin problem of PTS during a small break loss of coolant transient has been analyzed as a part of the international comparative assessment study. The allowable maximum reference nil-ductility transition temperatures are determined for various crack sizes.

• Proceedings of the KIEE Conference
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• 2003.07b
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• pp.1303-1305
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• 2003

It is very important to ensure system safety during the process of developing a system. Railway system is also devoting a great portion for the safety. Nowadays many countries leading railway industry have their own system assessment principles according to the situation of their train control systems. In this paper, several principles to derive Safety Integrity Level (SIL) are represented in the railway signalling system. The characteristics of those principles are also considered respectively.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.953-954
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• 2005

• The Journal of Engineering Geology
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• v.28 no.1
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• pp.1-9
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• 2018

We propose a borehole test technique to estimate permeability of rocks in borehole. The borehole tests are hydraulic injection tests such as leak-off test and hydraulic fracturing tests, which are originally conducted for stress or casing integrity assessment and not for permeability measurement. We use one-dimensional radial diffusion equation to interpret fluid injection test results in terms of permeability. We apply this technique to a leak-off test conducted at a depth of 700 m in a wellbore, where rock formation is mudstone. The estimated permeability is at an order of $10^{-16}m^2$, which is somewhat high but within the range reported for mudstones previously. Quantitative rick assessment suggests that an accurate measurement of open hole section length is important to improve reliability of results. More data may be needed to ensure the reliability of this technique. If validated, however, this technique can provide cost-effective estimation of in situ permeability without conducting independent permeability tests in borehole.

• Proceedings of the KWS Conference
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• 2006.10a
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• pp.15-17
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• 2006

Weld residual stresses can be a major concern in structural integrity assessments such as a nuclear power plant. In this paper, detailed weld residual stress analyses were presented for a typical multi-pass weld of pipe-butt weld and plate T-butt weld. The calculated residual stress distributions were compared with those of the measured data and recommended profiles in R6 and BS7910. Defect assessment which is based on the stress intensity factor(SIF) calculations was carried out for a plate T-butt weld with cracks considering the weld residual stress distributions.

• International Journal of Control, Automation, and Systems
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• v.4 no.1
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• pp.105-112
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• 2006

The main function of a reactor protection system is to maintain the reactor core integrity and the reactor coolant system pressure boundary. Generally, the reactor protection system adopts the 2-out-of-m redundant architecture to assure a reliable operation. This paper describes the safety assessment of a digital reactor protection system using the fault tree analysis technique. The fault tree technique can be expressed in terms of combinations of the basic event failures such as the random hardware failures, common cause failures, operator errors, and the fault tolerance mechanisms implemented in the reactor protection system. In this paper, a prediction method of the hardware failure rate is suggested for a digital reactor protection system, and applied to the reactor protection system being developed in Korea to identify design weak points from a safety point of view.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.21 no.8
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• pp.708-713
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• 2011

U.S. nuclear regulatory commission(NRC) regulatory guide(RG) 1.20 requires a comprehensive vibration assessment program(CVAP) for use in verifying the structural integrity of reactor vessel internals(RVI) for flow-induced vibrations prior to commercial operation. The CVAP program consist of vibration and fatigue analysis, a vibration measurement program, an inspection program, and a correlation of their results. One of the main purposes of the analysis program is to select measurement locations, however measurement locations can not be determined by only analysis results, therefore we developed selection criteria of measurement locations for advanced power reactor 1400(APR1400) RVI CVAP, It will be used to select measurement locations and instrument types for APR1400 RVI CVAP.

• Journal of the Korean Society for Railway
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• v.9 no.3 s.34
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• pp.264-270
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• 2006

In this paper we present a safety plan to be applied to the development of the TCS(Train Control System). The safety plan that can be applied to the life cycle of a system, from the conceptual design to the dismantlement, shows the whole process of the paper work in detail through the establishment of a goal, analysis and assessment, the verification. In this paper we study about the making a plan, the preliminary hazard analysis, the hazard identification and analysis to guarantee the safety of the TCS. The process far the verification of the system safety is divided into several steps based on the target system and the approaching method. The guarantee of the system safety and the improvement of the system reliability is fellowed by the recommendation of the international standards.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.161-166
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• 2008

In general structures, cleavage fracture may develop under the low constraint condition of larger scale yielding with a shallow surface crack. However, standard procedures for fracture toughness testing require very severe restrictions of specimen geometry. So the standard fracture toughness data makes the integrity assessment irrationally conservative. In this paper, cleavage fracture toughness tests have been made on side-grooved PCVN (precracked charpy V-notch) type specimens (10 by 10 by 55 mm) with varying crack depth, The constraint effects on the crack depth ratios are quantitatively evaluated by scaling model and Weibull stress method using 3-D finite clement method, After correction of constraint loss due to shallow crack depths, the statistical size effect are also corrected according to the standard ASTM E 1921 procedure, The results snowed a good agreement in the geometry correction regardless of the crack size, while some over-corrections were observed in the corrected values of $T_0$.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.333-338
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• 2001

The object of this work is to develop an assessment system for pipeline integrity. The system consists of four module applications for internal algorithm; the effect of corrosion in pipeline, crack, stress corrosion crack (SCC) and fatigue modules and the effect of cavity. Presently, the module of the external corrosion has been developed and the internal algorithm for the effect of corrosion in pipeline and the database of the system are described in this paper. The database of the system is separated to mainly four parts; geometry of pipeline, material properties, boundary conditions and general properties. Each components of the system are designed by user-friendly concept. This system may give a guideline for maintenance and modifications for the pipeline at the industrial sight. Furthermore, a procedure to evaluate an inspection interval is also provided.