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

• Nuclear Engineering and Technology
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• v.45 no.4
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• pp.439-458
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• 2013

REX-10 is a fully-passive small modular reactor in which the coolant flow is driven by natural circulation, the RCS is pressurized by a steam-gas pressurizer, and the decay heat is removed by the PRHRS. To confirm design decisions and analyze the transient responses of an integral PWR such as REX-10, a thermal-hydraulic system code named TAPINS (Thermal-hydraulic Analysis Program for INtegral reactor System) is developed in this study. Based on a one-dimensional four-equation drift-flux model, TAPINS incorporates mathematical models for the core, the helical-coil steam generator, and the steam-gas pressurizer. The system of difference equations derived from the semi-implicit finite-difference scheme is numerically solved by the Newton Block Gauss Seidel (NBGS) method. TAPINS is characterized by applicability to transients with non-equilibrium effects, better prediction of the transient behavior of a pressurizer containing non-condensable gas, and code assessment by using the experimental data from the autonomous integral effect tests in the RTF (REX-10 Test Facility). Details on the hydrodynamic models as well as a part of validation results that reveal the features of TAPINS are presented in this paper.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3732-3744
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• 2022

To investigate the effects of ocean conditions on the thermal stress and deformation caused by thermal stratification of a pressurizer surge line in a floating nuclear power plant (FNPP), the finite element simulation platform ANSYS Workbench is utilized to conduct the fluid-solid-thermal coupling transient analysis of the surge line under normal "wave-out" condition (no motion) and under ocean conditions (rolling and pitching), generating the transient response characteristics of temperature distribution, thermal stress and thermal deformation inside the surge line. By comparing the calculated results for the three motion conditions, it is found that ocean conditions can significantly improve the thermal stratification phenomenon within the surge line, but may also result in periodic oscillations in the temperature, thermal stress, and thermal deformation of the surge line. Parts of the surge line that are more susceptible to thermal fatigue damage or failure are determined. According to calculation results, the improvements are recommended for pipeline structure to reduce the effects of thermal oscillation caused by ocean conditions. The analysis method used in this study is beneficial for designing and optimizing the pipeline structure of a floating nuclear power plant, as well as for increasing its safety.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.132.1-132
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.74-74
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.657-662
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• 1996

Venting capability to prevent excess pressurization caused by loss of Residual Heat Removal System (RHRS) during mid-loop operation hat been evaluated analytically and the peak Reactor Coolant System (RCS) pressure was compared with the results of the MIDLOOP computer code. Even though analytical method if relatively simple, the results are in a good agreement with those of the computer code. For both methods, the peak pressures have not, exceeded the nozzle dam design pressure, if the vent paths such as pressurizer safety valves or a pressurizer manway are available in a closed RCS configuration with the nozzle dam installed.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.33 no.10
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• pp.1151-1157
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• 2009

Nuclear power plants applying for the continued operation over design life are required to address the effects of reactor water environment in fatigue design requirement of the ASME Code. Reactor water environmental effects are generally evaluated by calculating fatigue correction factors on fatigue usage. This paper describes the application for pressurizer surge line of environmental fatigue correction factors and the strain rate impact in the application. From this paper, the environmental fatigue correction factors resulted from the assumption of a step change in temperature are especially compared with those calculated from the data measured during plant startup. As a conclusion of this paper, the design transient conditions applied to the fatigue design may be conservative in case of the environmental fatigue evaluation.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.1
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• pp.19-24
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• 2012

This paper presents the results of finite element analysis for the effects of hydrostatic test and operating condition on the weld residual stress at dissimilar metal weld of a pressurizer safety nozzle in a nuclear power plant. For the study, the weld residual stress at ambient condition was analyzed using ABAQUS in the first place. After the weld residual stress analysis, the hydrostatic test condition and operating condition was applied to the same model one after another. The weld residual stress was observed to change due to the successive hydrostatic test and operating condition. The axial residual stresses on inner surface of the dissimilar metal weld and HAZ region were decreased by hydrostatic test and operating condition, which gives beneficial effect on preventing primary water stress corrosion cracking.

• Transactions of the Korean Society of Mechanical Engineers
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• v.19 no.8
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• pp.2064-2070
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• 1995

Characteristics of thermal flow stratification were studied experimentally by using the small scale pressurizer-surge line model. Thermal flow stratifications in the horizontal section of surge line were analyzed by the relation between the maximum temperature difference at any cross section in surge line and the Froude number representing the boundary conditions, i.e., in/out surge flow velocity and temperature difference of system. Thermal flow stratifications in outsurge flow decreased inversely proportional to the Froude number and did not exist for Fr>1. In insurge flow thermal flow stratifications disappeared near Fr=1.5, but resulted in the higher temperature difference than the case of outsurge flow.

• Nuclear Engineering and Technology
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• v.20 no.1
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• pp.1-8
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• 1988

During power plant operation, the flow rate from the CVCS makeup system is estimated using the continuity equation and mass balance equation, when the primary loop boron concentration change is required due to the power transient. For this purpose, primary loop, pressurizer and VCT(volume control tank)(in CVCS) are modeled by three control volumes which contain each mass and boron concentration. Connecting pipes between primary loop, pressurizer and CVCS are also modeled by time delay. Calculation for 14-2-6-2 (power 100-50-100) load follow case (at EOL, for KNU 7) is made using these models.