• Proceedings of the Korean Society of Precision Engineering Conference
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• 2002.05a
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• pp.474-477
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• 2002

Safety is very important to operate nuclear power plant. To guarantee the safety, nuclear power plant should be run without trouble. This paper presents the application of a failure diagnosis approach based on discrete event system theory to the pressurizer pressure control system for Pressurized Water Reactor. Also, this paper shows a scheme of failure diagnosis by distributed diagnoser.

• Fire Science and Engineering
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• v.27 no.5
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• pp.26-31
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• 2013

In nuclear power plant, anchor bolts for pressurizer supports are sufficiently used in terms of safety reason, but field inspections have reported that some bolts exceed the limit of their allowable hardness. Because the high level of hardness may lead to failures due to the stress corrosion or fracture toughness, a regular inspection is required for the bolts in nuclear power plant. Thus, this research measures the hardness of bolts currently used in pressurizer supports and then estimates maximum allowable stresses preventing failures by stress corrosion and fracture toughness. Using the ANSYS program, the stresses of the bolts in the regular condition and accidental condition have been calculated, and the possible maximum stress has been compared with the estimated allowable stresses. From the results, the stresses of bolts in the accidental condition satisfy the allowable safety stress from the stress corrosion failure. However, in the future, it shall be needed to consider the reflection of the structure assembling method on the assembling procedure to ensure the pressurizer integrity during maintenance period time.

• Nuclear Engineering and Technology
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• v.30 no.3
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• pp.194-201
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• 1998

DYLAM (Dynamic Logical Analytical Methodology) and its related methodologies are reviewed and found to have many favorable characteristics. Previous studies have shown that the DYLAM methodology represents an appropriate tool to study dynamic analysis. A hybrid model which is a synthesis of the DYLAM model, a system thermodynamic simulation model and a neural network predicative model, is implemented and used to analyze dynamically the CANDU pressurizer system. This study demonstrates that the hybrid model for system reliability analyses is effective.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.21 no.5
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• pp.670-678
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• 1997

A method to mitigate the thermal stratification phenomenon of pressurizer surge line is proposed by heating bottom outside of horizontal pipe. Unsteady two dimensional model has been used to numerically investigate an effect of heating the bottom of pipe. The dimensionless governing equations are solved by using the control volume formulation and SIMPLE algorithm. Temperature and streamline profiles of fluids and pipe walls with time are compared with the previous study result. The numerical result of this study shows that the outside heating can relaxate the thermal stratification flow of the pressurizer surge line. Maximum dimensionless temperature difference between hot and cold sections of the pipe inner wall which causes thermal stratification was reduced from 0.514 to 0.424 at dimensionless time 1, 632 and 1, 500 respectively.

• 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.44 no.3
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• pp.343-354
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• 2012

Primary water stress corrosion cracking (PWSCC) is a major safety concern in the nuclear power industry worldwide. PWSCC is known to initiate only in the condition in which sufficiently high tensile stress is applied to alloy 600 tube material or alloy 82/182 weld material in pressurized water reactor operating environments. However, it is still uncertain how much tensile stress is re-quired to generate PWSCC or what causes such high tensile stress. This study was performed to pre-dict the magnitude of weld residual stress and operating stress and compare it with previous experi-mental results for PWSCC initiation. For the study, a pressurizer safety nozzle was selected because it is reported to be vulnerable to PWSCC in overseas plants. The assessment was conducted by nu-merical analysis. Before performing stress analysis for plant conditions, a preliminary mock-up ana-lysis was done. The result of the preliminary analysis was validated by residual stress measurement in the mock-up. After verification of the analysis methodology, an analysis under plant conditions was conducted. The analysis results show that the stress level is not high enough to initiate PWSCC. If a plant is properly welded and operated, PWSCC is not likely to occur in the pressurizer safety nozzle.

• Nuclear Engineering and Technology
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• v.17 no.4
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• pp.302-316
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• 1985

Small break LOCA caused by a stuck-open PORV is one of the important contributors to nuclear power plant risk. This paper deals with the design of a pressurizer surveillance system using microcomputer to prevent the malfunction of system and has assessed the effect of this improvement through Probabilistic Risk Assessment (PRA) method. Micro-computer diagnoses the malfunction of system by a process checking method and performs automatically backup action related to each malfunction. Owing to this improvement, we can correctly diagnose “Spurious Opening”, “Fail to Reclose” and “Small break LOCA” which are difficult for operator to diagnose quickly and correctly and reduce the probability of a human error by an automatic backup action.

• 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.

• Journal of Welding and Joining
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• v.33 no.1
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• pp.87-93
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• 2015

This study was performed to develop repair technology for J-Groove dissimilar metal weld of pressurizer heater sleeve in nuclear power plants. Pad, J-Groove automatic welding and mechanical machining equipments to develop repair technology using 'Half Nozzle Repair' were designed and manufactured. To obtain the optimum welding process parameters during Pad temperbead overlay welding, several welding experiments using Taguchi method were conducted. Weldability of Pad overlay weld specimens was estimated by PT/RT test, FE-SEM, EDS and Vickers hardness test. Also, J-Groove welding to adjust weld shape conditions requiring in ASME Code was carried out and its integrity of weld specimens was evaluated through PT/RT test and optical microscope. Consequently, it was revealed that Pad and J-Groove overlay welding for dissimilar metal weld of pressurizer heater sleeve could be possible to meet Code standard without weld defect.