[en] The moisture separator reheaters (MSRs) at KEPCO's Kori 2 nuclear power station have been totally reconstructed on site and within existing MSR shells to eliminate severe operating problems and improve operating efficiency. The retrofit took only three weeks. (author)

[en] The regulatory Level-1 internal probabilistic safety assessment (PSA) model (MPAS, multi purpose analysis safety) is developed to apply to risk informed regulation. This MPAS model could be applied to develop the index for a graded regulation and to perform the independent risk analysis of Kori units 2 in KINS (Korea Institute of Nuclear Safety)

[en] The probabilistic safety assessment (PSA) results for Kori Unit 2 showed that the common cause failure (CCF) events of emergency service water system (ESWS) pump failure to run were identified as one of the dominant contributors to its internal event core damage frequency (CDF). The generic values of the CCF parameters were used in PSA projects for the Kori Unit 2. Thus, we performed the plant specific detailed CCF analysis to estimate the CCF parameters of ESWS pump failure to run for Kori Unit 2 with the CAFE-PSA(common CAuse Failure Event analysis program for PSA), a program to analyze CCF events in the ICDE database

[en] Korea Electric Power Research Institute is performing the relaxation of SG tube plugging(SGTP) level from 5% to 15%. The purpose of this study is to provide a discussion of the operating margin to the various Reactor Trip and Engineered Safety Features (ESF) actuation setpoints during steady-state and normal (Condition I) operating transients at the 15% SGTP conditions. Three Condition I transients, such as, 50% Step Load Rejection from 100% power, 10% Step Load Increase from 90% power and 5%/minute Ramp Load Increase from 15% to 100% power were analyzed. It is concluded that sufficient margin exists to various reactor trips and ESF actuation setpoints for the relaxed SGTP

[en] Existing methods did not take into consideration the individual characteristics of the systems and piping present in the flood areas. A quantitative analysis was selected only using the frequency of the flood incident reported to NPE for auxiliary buildings, turbine buildings, etc. In this paper, we performed a preliminary quantitative screening analysis for the systems and piping in the flood areas of the Kori Unit 2 Nuclear Power Plant (herein called 'Kori Unit 2') of Westinghouse design by applying the methodology described in EPRI 300200079, Pipe Rupture Frequencies for Internal Flooding Probabilistic Risk Assessments. 2013. A preliminary quantitative screening analysis of internal flooding PSA is performed according to the methodology of EPRI. Switching from using the NPE database to the methodology in EPRI, it is expected more areas will be added to the list of detailed analysis areas. The scenario CDF, which influences area CDF, increases more as flood frequency is increased by using EPRI instead of using the database of NPE

[en] Operator guideline of RHR loss while it is operating at mid-loop condition was developed through the validation performed in Kori Unit 2 simulator in March 2007. To demonstrate the effectiveness of operator action of the procedure, the validation processes were performed by methods including discussion method and the Kori Unit 2 simulator test. This validation concluded that for RHR loss during mid loop operation, the procedure was effective in restoring the plant to a safe, stable condition. Findings from the validation processes were incorporated in the procedure and guideline

[en] In this work, the Fuzzy Signed Digraph(FSD) method which has been researched for the fault diagnosis of industrial process plant systems is improved and applied to the fault diagnosis of the Kori-2 nuclear power plant pressurizer. A method for spurious faults elimination is also suggested and applied to the fault diagnosis. By using these methods, we could diagnose the multi-faults of the pressurizer and could also eliminate the spurious faults of the pressurizer caused by other subsystems. Besides the multi-fault diagnosis and system-wide diagnosis capabilities, the proposed method has many merits such as real-time diagnosis capability, independency of fault pattern, direct use of sensor values, and transparency of the fault propagation to the operators

[en] Highlights: • This study extends the equipment classification of a nuclear power plant based on the ISO 15926 standard. • This is followed by constructing a user-defined reference data server for the extended classification. • A method to exchange equipment specification data by using iRINGTools is also proposed. • The reference data server and iRINGTools are used to exchange equipment specification data of the nuclear power plant. - Abstract: In a nuclear power plant, classification is a key tool to manage equipment and materials. There could be different classifications corresponding to the different lifecycle phases and the challenge is to achieve and maintain interoperability, continuity, traceability between them. Therefore, it is important to define the equipment and materials classification based on standardized information models in order to effectively share and integrate lifecycle data of equipment and materials. To accomplish this, a method for standardized exchange of plant equipment and materials data based on ISO 15926 methodology in nuclear power plants is suggested. In the proposed method, we use an information model to expand the equipment and materials classification used for the operation and maintenance phase of a nuclear power plant. This is based on the ISO 15926 standard after analyzing the master data from the Shin Kori unit 1 and 2 of Korea Hydro & Nuclear Power Co., Ltd. (KHNP). Furthermore, as a data exchange tool in the proposed method, this study develops a method to exchange equipment and materials specification data by using iRINGTools.

[en] The new 16X16 fuel assembly design has been developed to improve the nuclear, thermal-hydraulic, and mechanical performance of the current 16X16 fuel assembly design in Kori 2. This study contains the static buckling analysis for the plate and cell models to evaluate the buckling characteristic of spacer grid. The static buckling strength of 16X16 mid grid design was estimated using the static buckling analysis result with cell models and the estimated static buckling strength was compared with the static buckling test result. The dynamic buckling strength of 16X16 mid grid design was estimated using previous test result for similar design. The estimated static buckling strength and dynamic buckling strength are well coincident with the test result of the improved 16X16 spacer grid

[en] The 84m discharge line of moisture separator reheater (MSR) in Kori-2 had been suffered by excessive vibration form the beginning of power generation. In 1993, a complementary measure was taken to reduce the vibration level by adding several supports in the MSR pipeline. However, the piping vibration was beyond the allowable limit and an appropriate countermeasure was required to prevent the fatigue failure of the pipeline from the abnormal vibration. In this study, the vibrational characteristics of MSR pipeline and the countermeasure for abnormal vibration was investigated. Among the several vibration reduction methods, piping layout change by smoothing the pipeline was applied to the MSR pipeline in Kori-2. Applying the countermeasure, the vibration level was found to dramatically reduce from 17cm/sec to 1.0cm/sec at the full-power operation condition