• Proceedings of the Korean Nuclear Society Conference
• /
• 2003.10a
• /
• pp.233.2-233.2
• /
• 2003

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.7 no.2
• /
• pp.21-25
• /
• 2011

Although many defects in PZR heater sleeve have been reported continually from operating experiences in oversea nuclear power plant, utilities get into difficulties in finding appropriate methods for diagnostics of the components due to the limited access or high radiation problems. Recently, as an alternative, diagnostics using Guided Wave Testing(GWT) are proposed and the attention of the methods has been growing gradually because of their long range inspection capability. This study is to investigate the effectiveness of GWT to detect PWSCC in welding points of PZR heater sleeve. Moreover, mode sensitivity analysis of GWT and optimal frequency for the diagnostics of PWSCC are presented by testing the mock-ups specimens that contain artificial flaws.

• Proceedings of the KSME Conference
• /
• 2005.11a
• /
• pp.104.1-104.1
• /
• 2005

• Proceedings of the KSME Conference
• /
• 2008.11a
• /
• pp.287-292
• /
• 2008

The applicability of stainless steel type 316N to the PZR surge-lines of OPR1000 and APR1400 is investigated. So far, strainless steel type 347 has been used for the OPR1000 surge-lines. The degree of improvement in the leak-before-break(LBB) and component design margin is evaluated when stainless steel type 347 is substituted by type 316N. For the study, the tensile and J-R tests on type 316N and type 347 stainless steels were performed at 316 and the microstructure of both types was examined. Stainless steel type 316N shows the higher values on the stress-strain curves, J-R curves and stress intensity, Sm, compared to those of type 347. Therefore, stainless steel type 316N ensures the higher LBB and component design margins. As a result, this study shows that stainless steel type 316N could substitute type 347 for the surge-lines of OPR1000 and APR1400.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2004.10a
• /
• pp.855-856
• /
• 2004

• Proceedings of the Korean Nuclear Society Conference
• /
• 2005.05a
• /
• pp.17-18
• /
• 2005

• Journal of Energy Engineering
• /
• v.25 no.1
• /
• pp.185-191
• /
• 2016

The Hybrid Safety Injection Tank is a passive safety injection system that enables the safety injection water to be injected into the reactor pressure vessel throughout all operating pressures by connecting the top of the SIT and the pressurizer(PZR). In this study, the condition for balancing the pressure between the Hybrid SIT and PZR was derived theoretically. The pressure balancing condition was set at the point where the velocity of the Hybrid SIT coolant injected into the Direct Vessel Injection(DVI) line was at or above zero. If the condition was derived from a pressure network for the Hybrid SIT, pressurizer, and reactor pressure vessel, the pressure difference between the pressurizer and SIT is less than 0.07 MPa.

• Journal of Welding and Joining
• /
• v.27 no.4
• /
• pp.32-37
• /
• 2009

Alloy 600 and Alloy 82/182 materials have been used widely in PWR plants. But these materials are known to be susceptible to PWSCC(Primary Water Stress Corrosion Cracking). Recently, there have been several PWSCC events in major components due to repair welding, because repair welding in the dissimilar metal welds during the construction increases residual stress significantly on the inner surface of welds. In this paper, various residual stress analyses for repair welding were performed using FEM to check the effect of repair welding on residual stress distributions in PZR safety/relief nozzle. The results indicate that for inside surface repair welding, high tensile residual stress is developed on the inside surface of the nozzles.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.14 no.2
• /
• pp.16-23
• /
• 2018

In order to investigate the ways to enhance the seismic performance of APR1400 seismic fragile equipment by direct design changes, four equipment such as Reactor Vessel Support, Integrated Head Assembly, Remote Shutdown Console, and Pressurizer are reviewed using information of the main dimensions, seismic stress evaluation results, design FRS, etc. in this paper. In addition to the direct reinforcement of equipments, the feasibility of seismic isolation for the safety related cabinet is also investigated and the actual adaption plan of a commercial spring-damper system is briefly reviewed.

• Nuclear Engineering and Technology
• /
• v.45 no.2
• /
• pp.179-190
• /
• 2013

A station blackout experiment called SBO-01 was performed at the ATLAS facility. From the SBO-01 test, the station blackout scenario can be characterized into two typical phases: A first phase characterized by decay heat removal through secondary safety valves until the SG dryouts, and a second phase characterized by an energy release through a blowdown of the primary system after the SG dryouts. During the second phase, some physical phenomena of the change over a pressurizer function, i.e., the pressurizer being full before the POSRV $1^{st}$ opening and then its function being taken by the RV, and the termination of normal natural circulation flow were identified. Finally, a core heatup occurred at a low core water level, although under a significant amount of PZR inventory, whose drainage seemed to be hindered owing to the pressurizer function by the RV. The transient of SBO-01 is well reproduced in the calculation using the MARS code.