• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.474.1-474
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.149-149
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.643-644
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.417-422
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• 1998

he Reactor Coolant Gas Vent System(RCGVS) design for Ulchin Nuclear Power Plant Units 3&4(UCN 3&4) has been improved from the Yonggwang Nuclear Power Plant Units 3&4(YGN 3&4) based on the evaluation results for depressurization capability tests performed at YGN 3&4. There has been a series of plant safety analyses for Natural Circulation Cooldown(NCC) event and thermo-dynamic analyses with RELAP5 code for the steam blowdown Phenomena in order to optimize the orifice size of UCN 3&4 RCGVS. Baesd on these analyses results, the RCGVS orifice size for UCN 3&4 has been reduced to 9/32 inch from the l1/32 inch for YGN 3&4. The depressurization capability tests, which were performed at UCN 3 in order to verify the FSAR NCC analysis results, show that the RCGVS depressurization rates are being within the acceptable ranges. Therefore, it is concluded that the orificed flow path of UCN 3&4 RCGVS is adequately designed, and can provide the safety-grade depressurization capability required for a safe plant operation.

• Journal of the Korea Safety Management & Science
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• v.7 no.4
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• pp.141-163
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• 2005

The international SPICE (Software Process Improvement and Capability determination) Project ISO/IEC 15504(SPICE : Software Process Improvement and Capability determination) is an emerging International Standard on SPA(Software Process Assessment). A prime motivation for developing this standard has been the perceived need for an internationally recognized software process assessment framework that pulls together the existing public and proprietary models and methods. A SPICE assessment can be considered as one of representative SPA model since assessors assign ratings to indicators and metrics to measure the capability of software process. But this models doesn't provide a systematic measurement procedures and dynamic method for SPI(Software Process Improvement). Through the evaluation of SPICE is capable of providing a substantiated basis for using the notion of capability, as well as providing information for nacessary improvements to the standard using 6sigma process. As a result, this paper propose a measurement procedure and guidelines for application of 6sigma process to guarantee the reliability in SPI and suggest the structure to support SPI on overall organization.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2812-2820
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• 2020

House load operation (HLO) occurs when the generator supplies power to the house load without triggering reactor trips during grid disturbances. In Korea, the HLO capability of optimized power reactor 1000 (OPR1000) plants has prevented several reactor trips. Operational experiences demonstrate the difference in the reactor trip incidence due to grid disturbances between OPR1000 plants and Westinghouse plants in Korea, attributable to the availability of the HLO capability. However, probabilistic safety assessments (PSAs) for OPR1000 plants have not considered their specific design features in the initiating event analyses. In an at-power PSA, the HLO capability can affect the initiating event frequencies of general transients (GTRN) and loss of offsite power (LOOP), resulting from transients within the grid system. The initiating event frequencies of GTRN and LOOP for an OPR1000 plant are reduced by 17.7% and 78.7%, respectively, compared to the Korean industry-average initiating event frequencies, and its core damage frequency from internal events is reduced by 15.2%. The explicit consideration of the HLO capability in initiating event analyses makes significant changes in the risk contributions of the initiating events. Consequently, for more realistic at-power PSAs in Korea, we recommend incorporating plant-specific HLO-related design features when estimating initiating event frequencies.

• Journal of Applied Reliability
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• v.9 no.3
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• pp.165-176
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• 2009

The risk management case is an organization's formal arrangement to ensure the safety of its work activity within risk management system. It allows an organization to demonstrate its capability in achieving its safety objectives and in meeting regulatory requirements. This paper presents how the safety assessments are described, prepared and maintained to meet the criteria specified by the upcoming safety regulations. We propose the elements of risk management system that include arrangements for the ongoing identification of hazards, assessment of risks and the implementation of necessary control measures.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1277-1288
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• 2018

In this work, the reactor kinetics capability is used to compute the design safety parameters in a PWR due to complete loss of coolant flow during protected and unprotected accidents. A thermal-hydraulic code coupled with a point reactor kinetic model are used for these calculations; where kinetics parameters have been developed from the neutronic SRAC code to provide inputs to RELAP5-3D code to calculate parameters related to safety and guarantee that they meet the regulatory requirements. In RELAP5-3D the reactivity feedback is computed by both separable and tabular models. The results show the importance of the reactivity feedback on calculating the power which is the key parameter that controls the clad and fuel temperatures to maintain them below their melting point and therefore prevent core melt. In addition, extending modeling capability from separable to tabular model has nonremarkable influence on calculated safety parameters.

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.1045-1052
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• 2009

A calculation model is developed to predict the transient free surface flow on the containment floor following a loss-of-coolant accident (LOCA) of pressurized water reactors (PWR) for the use of debris transport evaluation. The model solves the two-dimensional Shallow Water Equation (SWE) using a finite volume method (FVM) with unstructured triangular meshes. The numerical scheme is based on a fully explicit predictor-corrector method to achieve a fast-running capability and numerical accuracy. The Harten-Lax-van Leer (HLL) scheme is used to reserve a shock-capturing capability in determining the convective flux term at the cell interface where the dry-to-wet changing proceeds. An experiment simulating a sudden break of a water reservoir with L-shape open channel is calculated for validation of the present model. It is shown that the present model agrees well with the experiment data, thus it can be justified for the free surface flow with accuracy. From the calculation of flow field over the simplified containment floor of APR1400, the important phenomena of free surface flow including propagations and interactions of waves generated by local water level distribution and reflection with a solid wall are found and the transient flow rates entering the Holdup Volume Tank (HVT) are obtained within a practical computational resource.

• The Transactions of the Korean Institute of Electrical Engineers A
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• v.53 no.3
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• pp.129-134
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• 2004

The Available Transfer Capability (ATC) is defined as the measure of the transfer capability remaining in the physical transmission network for further commercial activity above already committed uses. The ATC determination s related with Total Transfer Capability (TTC) and two reliability margins-Transmission Reliability Capability (TRM) and Capacity Benefit Margin(CBM) The TRM is the component of ATC that accounts for uncertainties and safety margins. Also the TRM is the amount of transmission capability necessary to ensure that the interconnected network is secure under a reasonable range of uncertainties in system conditions. The CBM is the translation of generator capacity reserve margin determined by the Load Serving Entities. This paper describes a method for determining the TTC and TRM to calculate the ATC in the Bulk power system (HL II). TTC and TRM are calculated using Power Transfer Distribution Factor (PTDF). PTDF is implemented to find generation quantifies without violating system security and to identify the most limiting facilities in determining the network’s TTC. Reactive power is also considered to more accurate TTC calculation. TRM is calculated by alternative cases. CBM is calculated by LOLE. This paper compares ATC and TRM using suggested PTDF with using CPF. The method is illustrated using the IEEE 24 bus RTS (MRTS) in case study.