• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.23 no.1
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• pp.49-55
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• 2013

A reactor vessel internals comprehensive vibration assessment program(RVI CVAP) of an advanced power reactor 1400(APR1400) is being performed as a non-prototype category-2 type of reactor based on the US nuclear regulatory commission regulatory guide(NRC RG) 1.20. The aim of this paper is to present the results of structural response analysis and measuring locations of a upper guide structure(UGS) assembly of the APR1400 reactor. The analysis results of the UGS assembly show that the specified integrity levels meet the design acceptance criteria. Also, the measuring locations are determined by the analysis results of the UGS assembly and selection criteria of previous study. These analysis results and measuring locations will be used as a guide to design a measurement system for the APR1400 RVI CVAP.

• Nuclear Engineering and Technology
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• v.45 no.2
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• pp.249-256
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• 2013

In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful preparation is essential to the measurement program, because data acquisition must be performed only once. The optimized design of a vibration and stress measurement system for the RVI CVAP is essential to verify the integrity of the APR1400 RVI. We successfully designed a vibration and stress measurement system for the APR1400 RVI CVAP based on the design materials, the hydraulic and structural analysis results, and performance tests of transducers in an extreme environment. The measurement system designed in this paper will be utilized for the APR1400 RVI CVAP as part of the first construction project in Korea.

• 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 of Safety
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• v.16 no.3
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• pp.125-134
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• 2001

The Advanced Pressurized rater Reactor 1400(APR1400) system is advanced of the successful Korean Nuclear Power Plants(KSNP) design which meets functional needs for safety enhancement reliability improvement, and control in the human-computer monitoring system. Therefore this paper describes the scoring model in order to justify the reliability and safety in APR 1400 under uncertainty. The structure of this paper consists of the human engineering, risk safety, quality function, safety organization management factors of the qualitative factors in chapter 2, and the expectation results of the normalized scoring model in chapter 3. Finally, the proposed reliability model have provided the technical flexibility not only for functional control fields but also for accidents protection systems in APR 1400 under uncertainty.

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.963-970
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• 2018

The U.S. Nuclear Regulatory Commission, Regulatory Guide 1.20 needs to perform a comprehensive vibration assessment program for reactor internals during preoperational and startup testing for nuclear power plants and extended power uprate. Although the measurement program is one of the core programs, it is rarely carried out except for a first-of-a-kind or a unique design. This article describes measurement results of vibration and stress for the comprehensive vibration assessment program for an APR-1400 reactor internals. The measurement was performed at an upper guide structure during the pre-core hot functional test of Shin Kori unit 4 reactor internals because the Shin Kori unit 3 and 4 are the first construction project for the APR-1400, and the upper guide structure assembly was to design change compared with the valid prototype. We confirmed that all measured results are within the test acceptance criteria. It means that the structural integrity of the APR-1400 reactor internals was secured for the flow-induced vibration.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.2047-2052
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• 2004

Inconel-600 alloy has been used as steam generator tube material for current pressurized water reactors (PWRs). The long-term operation of steam generators showed that the use of this material induced localized corrosion damages and increased tube wear of steam generator. To protect these problems, steam generator tube material is being changed to Inconel-690 alloy. Based on the current trend, we have chosen Inconel 690 as the Advanced Power Reactor 1400 (APR1400) steam generator(SG) tube material and performed the design optimization of preventive measure against tube fretting wear for the APR1400 steam generator. In this paper, we examined the technical consideration in this modification : the selection of material, wear characteristics, effect of the Egg-crate Flow Distribution Plate installation, and effect analysis of vertical strip installation.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2014.10a
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• pp.99-101
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• 2014

A Comprehensive Vibration Assessment Program (CVAP) for steam generator internals (SGI) of Advanced Power Reactor 1400 (APR1400) is being performed in accordance with the United States Nuclear Regulatory Commission (U.S. NRC) Regulatory Guide 1.20 (RG 1.20) revision 3. This paper studies the vibration characteristics of moisture separator assembly as part of the vibration and stress analysis program for APR1400 SGI CVAP. The natural frequencies, mode shapes, and structural behavior of moisture separator assembly were investigated through modal analysis using finite element method and experimental measurement. Since the moisture separator consists of several items with complicated shape, an idealized shell model was used in the finite element analysis. Group of local modes caused by moisture separators and significant modes of shroud and separator support plate were identified. The results of this paper are to be utilized in the structural response analysis of moisture separator assembly.

• Journal of Radiation Protection and Research
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• v.28 no.2
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• pp.127-135
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• 2003

The primary contributors to the total occupational radiation exposure in operating nuclear power plants are operation and maintenance activities doting refueling outages. The Advanced Power Reactor 1400 (APR1400) includes a number of design improvements and plans to utilize advanced maintenance methods and robotics to minimize the annual collective dose. The major radiation exposure reduction features implemented in APR1400 are a permanent refueling pool seal, quick opening transfer tube blind flange, improved hydrogen peroxide injection at shutdown, improved permanent steam generator work platforms, and more effective temporary shielding. The estimated average annual occupational radiation exposure for APR1400 based on the reference plant experience and an engineering judgment is determined to be in the order of 0.4 man-Sv, which is well within the design goal of 1 man-Sv. The basis of this average annual occupational radiation exposure estimation is an eighteen (18) month fuel cycle with maintenance performed to steam generators and reactor coolant pumps during refueling outage. The outage duration is assumed to be 28 days. The outage work is to be performed on a 24 hour per day basis, seven (7) days a week with overlapping twelve (12) hour work shifts. The occupational radiation exposure for APR1400 is also determined by an alternate method which consists of estimating radiation exposures expected for the major activities during the refueling outage. The major outage activities that cause the majority of the total radiation exposure during refueling outage such as fuel handling, reactor coolant pump maintenance, steam generator inspection and maintenance, reactor vessel head area maintenance, decontamination, and ICI & instrumentation maintenance activities are evaluated at a task level. The calculated value using this method is in close agreement with the value of 0.4 man-Sv, that has been determined based on the experience aid engineering judgement. Therefore, with the As Low As Reasonably Achievable (ALARA) advanced design features incorporated in the design, APR1400 design is to meet its design goal with sufficient margin, that is, more than a factor of two (2), if operated on art eighteen (18) month fuel cycle.

• Journal of Energy Engineering
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• v.17 no.2
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• pp.67-76
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• 2008

A thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been constructed at KAERI (Korea Atomic Energy Research Institute). Recently several integral effect tests for the reflood period of a LBLOCA (Large Break LOss of Coolant Accident) of the APR1400 have been performed with the ATLAS. In the APR1400 a high flow condition is changed to a low flow condition due to an fluidic device during an operation of the SIT. As the self-controlled fluidic device was not installed in the ATLAS, a set of characterization tests was performed to simulate its injection capability from the SIT for the APR1400 simulation. In the ATLAS the required SIT flow rate in the high flow condition was acquired by installing orifices with an optimized flow area to throttle the SIT discharge line and the low flow condition was achieved by changing the opening of the flow control valve in the SIT injection line. The test results showed that the safety injection systems of the ATLAS could simulate the required high and low flow rates of the SIT for the APR1400 simulation efficiently.

• Nuclear Engineering and Technology
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• v.36 no.1
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• pp.24-35
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• 2004

In order to analyze the hydrogen distribution during a severe accident in the APR1400 containment, GASFLOW II was used. For the APR1400 NPP, a hydrogen mitigation system is considered from the design stage, but a fully time-dependent, three-dimensional analysis has not been performed yet. In this study GASFLOW code II is used for the three-dimensional analysis. The first step to analysis involving hydrogen behavior in a full containment with the GASLOW code is to generate a realistic geometry model, which includes nodalization and modeling of the internal structures such as walls, ceilings and equipment. Geometry modeling of the APR1400 is conducted using GUI program by overlapping the containment cut drawings in a graphical file format on the mesh view. The total number of mesh cells generated is 49,476. And the calculated free volume of the APR1400 containment by GASFLOW is almost the same as the value from the GOTHIC modeling. A hypothetical SB-LOCA scenario beyond design base accident was selected to analyze the hydrogen behavior with the hydrogen mitigation system. The source of hydrogen and steam for the GASFLOW II analysis is obtained from a MAAP calculation. Combustion pressure and temperature load possibilities within the compartments used in the GOTHIC analysis are studied based on the Sigma-Lambda criteria. Finally the effectiveness of HMS installed in the APR1400 containment is evaluated from the point of severe accident management