• 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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• 2003.11a
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• pp.84-91
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• 2003

Flow-Induced Vibration of steam generator tubes may result in fretting wear damage at the tube-to-support locations. KSNP(Korean Standard Nuclear Power plant) steam generators experienced fretting wear in the upper part of U-bend above the central cavity region of steam generators. This region has conditions susceptible to the flow-induced vibration, such as high flow velocity, high void fraction, and longer unsupported span. To improve its performance, APR1400 steam generator is designed with additional supports in this region to reduce unsupported span and to reduce peak velocity in the central cavity region. In this paper, we examined its performance improvement using ATHOS code. The thermal-hydraulic condition in the region of secondary side of APR1400 steam generator is obtained using the ATHOS3 code. The effective mass for modal analysis is calculated using the void fraction, enthalpy, and operating pressure information from ATHOS3 code result. With the effective mass distribution along the tube, natural frequency and mode shape is obtained using ANSYS code. Finally, stability ratios and real mean squared displacements for selected tubes of the APR1400 steam generator are computed. From these results, the current design of the APR1400 steam generator are examined.

• 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.

• Proceedings of the KSME Conference
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• 2003.04a
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• pp.49-54
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• 2003

Most of the operating pressurized water reactors (PWRs)has chosen Inconel 600 as steam generator tubing. The long-term operation of steam generators showed that the use of this material induced localized corrosion damages. The current trend is using Inconel 690 as a tube material for the replacement steam generators. Based on the current trend, we have chosen Inconel 690 for the advanced Power Reactor 1400 (APR1400) steam generator tube material. In this paper, we examined the technical consideration in this modification: the effect of chemical composition, thermal conductivity, corrosion resistance and wear characteristics

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.25 no.9
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• pp.599-605
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• 2015

The revised U.S. Nuclear Regulatory Commission(NRC), Regulatory Guide(RG) 1.20, rev.3 requires the evaluation of the potential adverse effects from pressure fluctuations and vibrations on piping and components for the reactor coolant, steam, feedwater, and condensate systems. Detailed vibration analyses for the systems attached to the steam generator are very difficult, because these piping systems are very complicated. This paper suggests a screening method for the flow-induced vibration of acoustic resonances and pump-induced vibration of the piping systems attached to the steam generator in order to conduct the APR1400 comprehensive vibration assessment program. This paper seeks to address the areas such as potential vibration sources, and methods to prevent the occurrence of acoustic resonances and pump-induced vibration of piping systems attached to the steam generator, for conducting the APR1400 comprehensive vibration assessment program. The screening method in this paper will be used to estimate the flow-induced vibration of the piping systems attached to the steam generator for the APR1400.

• 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.

• Nuclear Engineering and Technology
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• v.34 no.4
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• pp.358-369
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• 2002

Abundant information about analyses of single steam generator tube rupture (SGTR) events is available because of its importance in terms of safety. However, there are few literatures available on analyses of multiple steam generator tube rupture (MSGTR) events. In addition, knowledge of transients and consequences following a MSGTR event are very limited as there has been no occurrence of MSGTR event in the commercial operation of nuclear reactors. In this study, a postulated MSGTR event in an APR1400 is analyzed using thermal-hydraulic system code MARSI.4. The present study aims to examine the effects of affected steam generator selection. The main steam safety valve (MSSV) lift time for four cases are compared in order to examine how long operator response time is allowed depending on which steam generate. (S/G) is affected. The comparison shows that the cases where two steam generators are simultaneously affected allow longer time for operator action compared with the cases where a single steam generator is affected. Furthermore, the tube ruptures in the steam generator where a pressurizer is connected leads to the shortest operator response time.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2014.10a
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• pp.102-105
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• 2014

In accordance with U.S. Nuclear Regulatory Commission regulatory guide(NRC RG) 1.20(Rev.3), we are writing a comprehensive vibration assessment program(CVAP) design control document(DCD) and a technical report for U.S. NRC design certification(DC) of an Advanced Power Reactor 1400(APR1400) nuclear power plant(NPP). CVAP of an APR1400 NPP for U.S. NRC DC is classified as a non-prototype category 1 type. Therefore, CVAP DCD of reactor vessel internals(RVI) and steam generator internals(SGI) consist of analysis and full inspection program. However, piping system of primary and secondary system will be described as measurement program.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.1
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• pp.18-27
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• 2019

The Advanced Power Reactor 1400 (APR1400) Steam Generator (SG) uses alloy 690 as a tube material and SA-508 Grade 3 Class 1 as a tubesheet material to form tube-to-tubesheet joint through hydraulic expansion process. In this paper, the residual stresses in the SG tube-to-tubesheet contact area was investigated by applying Model-Based System Engineering (MBSE) methodology and the V-model. The use of MBSE transform system description into diagrams which clearly describe the logical interaction between functions hence minimizes the risk of ambiguity. A theoretical and Finite Element Methodology (FEM) was used to assess and compare the residual stresses in the tube-to-tubesheet contact area. Additionally, the axial strength of the tube to tubesheet joint based on the pull-out force against the contact joint force was evaluated and recommended optimum autofrettage pressure to minimize residual stresses in the transition zone given. A single U-tube hole and tubesheet with ligament thickness was taken as a single cylinder and plane strain condition was assumed. An iterative method was used in FEM simulation to find the limit autofrettage pressure at which pull-out force and contact force are of the same magnitude. The joint contact force was estimated to be 20 times more than the pull-out force and the limit autofrettage pressure was estimated to be 141.85MPa.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.343-347
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• 2001

Advanced Pressurized Reactor 1400(APR-1400), which is a standard evolutionary advanced light water reactor(ALWR), has been developed from 1992 as one of long-term Government Project(G-7). The APR-1400 is designed to operate at the rated output of 4000MWt to produce an electric power output of around 1450MWe. The balance of plant (BOP) for the secondary system consists of main steam, feedwater, condensate, turbine generator and auxiliary system. In this paper, we describe the major design features of secondary component, balance of plant configuration, and then the turbine cycle thermal performance evaluation using PEPSE code.