• The KSFM Journal of Fluid Machinery
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• v.17 no.6
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• pp.77-83
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• 2014

The vessel of integral reactor contains its major primary components such as the fuel and core, pumps, steam generators, and a pressurizer, so its size is proportional to the required space for the installation of each component. The steam generators take up the largest volume of internal space of reactor vessel and their volumes is substantial for the overall size of reactor vessel. Reduction of installation space for steam generators can lead to much smaller reactor vessel with resultant decrease of overall cost for the components and related facilities. A printed circuit heat exchanger is one of the compact types of heat exchangers available as an alternative to conventional shell and tube heat exchangers. Its name is derived from the procedure used to manufacture the flat metal plates that form the core of the heat exchanger, which is done by chemical milling. These plates are then stacked and diffusion bonded, converting the plates into a solid metal block containing precisely engineered fluid flow passages. The overall heat transfer area and pressure drops are evaluated for the steam generator based on the concept of the printed circuit heat exchanger in this study. As the printed circuit heat exchanger is known to have much larger heat transfer area density per unit volume, we can expect significantly reduced steam generator compared to former shell and tube type of steam generator. For the introduction of new steam generator, two design requirements are considered: flow area ratio between primary and secondary flow paths, and secondary side parallel channel flow oscillation. The results show that the overall volume of the steam generator can be significantly reduced with printed circuit type of steam generator.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.05a
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• pp.458-458
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.725-726
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• 2005

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.28-33
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• 2010

As a part of license renewal for the continued operation of Wolsong unit 1, the periodic safety review report was submitted near the end of design lifetime, 2012, and now is under reviewing. Major components of primary system such as pressure tubes, feeder pipes and so on are being replaced and many components of secondary system are also being repaired. So the license renewal of Wolsong unit 1 is expected to be acquired without significant issues. But on the other hand, steam generators of Wolsong unit 1 had the good performance and therefore the replacement and repair for the steam generator are not needed. Recently it is reported that some cracks were detected in a few of european steam generator with Alloy 800 tubes and the cause of cracks was the outer diameter stress corrosion cracking(ODSCC) due to the concentration of chemical impurities on the outer surface of tube. Accordingly the overall review on this issue was performed. The long-term operation is likely to results to form the concentration mechanism for the tube corrosion as the sludge build-up in the secondary side of steam generator and the crack in the crevice between tube and tube-sheet and expansion transitions is apt to be occurred. In this paper, the history of steam generator inspection and operation of Wolsong unit 1 are reviewed and the reliability of steam generator tube is evaluated and the steam generator aging management program for Wolsong unit 1 is introduced.

• Transactions of the Korean Society of Mechanical Engineers
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• v.17 no.1
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• pp.200-214
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• 1993

This paper presents the results of fluidelastic instability analysis performed for the U-tube bundle of a Westinghouse model 51 steam generator, one of the recirculating types designed at an early stage, in which the principal region of external cross-flow is associated with the U-bend portion of tube. The prerequisites for this analysis are detailed informations of the secondary side flow conditions in the steam generator and the free vibration behaviours of the U-tubes. In this study, the three-dimensional two-phase flow field in the steam generator has been calculated employing the ATHOS3 steam generator two-phase flow code and the ANSYS engineering analysis code has been used to calculate the free vibration responses of specific U tubes under consideration. The assessment of the potential instability for the suspect U-tubes, which is the final analysis process of the present work, has been accomplished by combining the secondary side velocity and density distributions obtained from the ATHOS3 prediction with the relative modal displacement and natural frequency data calculated using the ANSYS code. The damping of tubes in two-phase flow has been deduced from the existing experimental data by taking into account the secondary side void fraction effect. In operation of the steam generator, the tube support conditions at the tube-to-tube support plate intersections due to either tube denting degradation or deposition of tube support plate corrosion products or ingression of dregs. Thus, various hypothetical cases regarding the tube support conditions at the tube-to-tube support plate intersections have been considered to investigate the clamped support effects on the forced vibration response of the tube. Also, the effect of anti-vibration bars support in the curved portion of tube has been examined.

• 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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.16-19
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• 2010

In this study, the axial ODSCC occurrence of domestic OPR 1000 steam generator tube was caused by the tube weakness and the sludge accumulation in the secondary side of steam generator. Inconel 600 HTMA used as tube material is related to most of tube leakage accidents in the world and also these ODSCCs were detected mainly at the 5th TSP(Tube Support Plate) to the 8th TSP of hot leg side. These elevations(5th TSP to 8th TSP) pave the way for the sludge accumulation. As a result of EC(Eddy Current) Bobbin and RPC data analysis, ODSCCs were occurred at contact points of tube and tube support plate. The more accumulated sludge, the higher occurrence frequency of ODSCC.

• Journal of Magnetics
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• v.16 no.2
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• pp.88-91
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• 2011

Steam generator tubes (SGTs) in nuclear power plants (NPPs) are a boundary between the primary side generating heat by nuclear fission and the secondary side generating electric power by a turbine. The water inside the SGT is high temperature and high pressure. Therefore, defects and magnetic phases (MPs) are partly produced in non-magnetic SGT by high stresses and temperatures. This causes trouble regarding the safety of SGTs but it is difficult to detect the MP using the conventional eddy current technique (ECT). In particular, a circumferential defect (CD) and circumferential magnetic phase (CMP) cannot detected by ECT. Consequently, a new method is needed to detect CDs and CMPs in SGT. A new U-type yoke with two types of coils was designed and the reactance signal by the CMPs and CDs in the SGT material was simulated.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1999.11a
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• pp.125-130
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• 1999

Corrosion of steam generator tubes is the major issue affecting selection of secondary water chemistry parameters. The objective of secondary side water chemistry control is to minimize corrosion damage and to thereby maximize the reliability and economic performance of the secondary system. To achieve this objective, the water chemistry has to be compatible with all parts of the system including steam generators.(Omitted)

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2395-2406
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• 2023

To study the influence of parameter uncertainty in small pressurized water reactor (SPWR) once-through steam generator (OTSG), the nonlinear mathematical model of the SPWR is firstly established. Including the reactor core model, the OTSG model and the pressurizer model. Secondly, a control strategy that both the reactor core coolant average temperature and the secondary-side outlet pressure of the OTSG are constant is adopted. Then, the uncertainty quantification method is established based on Latin hypercube sampling and statistical method. On this basis, the quantitative platform for parameter uncertainty of the OTSG is developed. Finally, taking the uncertainty in primary-side flowrate of the OTSG as an example, the platform application work is carried out under the variable load in SPWR and step disturbance of secondary-side flowrate of the OTSG. The results show that the maximum uncertainty in the critical output parameters is acceptable for SPWR.