• The Korean Journal of Applied Statistics
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• v.29 no.7
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• pp.1385-1398
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• 2016

A pressurizer is a major equipment system in a nuclear power plant (NPP) and controls the reactor cooling system pressure within the allowable range. Faults in the pressurizer can be critical to the NPP; therefore, early fault detection in the pressurizer is significant for NPP safety. This study applies Mahalanobis Taguchi system (MTS), which is one of the promising pattern classification methods, based on the Mahalanobis distance concept and Taguchi quality engineering theory to the early fault detection problem of the pressurizer pressure control system. We conducted experiments using data from full scope NPP simulator based on a pressurizer pressure transmitter faults scenario to validate the faults detection performance of MTS. As a result, MTS can rapidly detect the faults compared to conventional faults detection based on single sensor monitoring.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1993.11a
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• pp.103-108
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• 1993

In nuclear power plants using pressurized water as the main coolant, it is necessary to maintain system pressure within operational range. During transients, the coolant shrinks and expands causing insurge and outsurge of coolant in the pressurizer. In CANDU system, the pressure is controlled mainly by the pressurizer/degasser-condenser system. In CANDU system, the control of heat transport system pressure is achieved by giving heat to the pressurizer by activating the heaters to compensate a diminution in pressure or by removing heat from the pressurizer by bleeding steam to the degasser-condenser to compensate an increase in pressure. This study aims at developing a theoretical model capable to simulate various operational transients in the CANDU primary heat transport system (PHTS), applicable to CANDU engineering simulator on real time basis.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.310-315
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• 1997

System-Integrated Modular Advanced Reactor (SMART), where major primary components such as modular helical steam generator and self regulating pressurizer are integrated into reactor vessel, is currently under development. The pressurizer is designed to control the primary pressure mainly with partial pressure of nitrogen gas and to maintain the fluid temperature as low as possible for the purpose of minimizing steam contribution. The steam generator (SG) is designed to produce super-heated steam inside tube at power operation. Because the in-vessel pressurizer and in-vessel SG are classified as the characteristic components of SMART, it is important to perform a steady state calculation of these components in order to evaluate the adoption of these components. A steady state analysis of the in-vessel pressurizer and in-vessel SG has been performed under normal power operation and the results show an acceptable performance of the components.

• Nuclear Engineering and Technology
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• v.45 no.3
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• pp.347-360
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• 2013

This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA) and the loss-of-feedwater accident (LOFW) in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF), a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

• 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.

• Journal of Korea Society of Digital Industry and Information Management
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• v.19 no.3
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• pp.25-38
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• 2023

Instrumentation and control systems measure and control various variables of nuclear facilities to operate nuclear power plants safely. A diverse protection system, a representative instrumentation and control system, generates a reactor trip and turbine trip signal by high pressure in a pressurizer and containment to satisfy the design requirements 10CFR50.62. Also, it generates an auxiliary feedwater actuation signal by low water levels in steam generators. Cybersecurity has become more critical as digital technology is gradually applied to solve problems such as performance degradation due to aging of analog equipment, increased maintenance costs, and product discontinuation. This paper analyzed possible cybersecurity threat scenarios in the diverse protection system using attack trees. Based on the analyzed cybersecurity threat scenario, we calculated the probability of attack occurrence and confirmed the cybersecurity risk in connection with the asset value.

• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.3
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• pp.389-396
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• 2018

The Kori-1 Nuclear Power Plant (NPP), WH 2-Loop Pressurized Water Reactor (PWR) operated for approximately 40 years in Korea, was permanently ceased on June 18, 2017. To reduce worker exposure to radiation by reducing the dose rate in the system before starting main decommissioning activities, the permanently ceased Kori-1 NPP will be subjected to full system decontamination. Generally, the range of system decontamination includes Reactor Pressure Vessels (RPV), Pressurizer (PZR), Steam Generators (SG), Chemical & Volume Control System (CVCS), Residual Heat Removal System (RHRS), and Reactor Coolant System (RCS) piping. In order to decontaminate these systems and equipment in an effective manner, it is necessary to evaluate the influence of the flow characteristics in the RCS during the decontamination period. There are various methods of providing circulating flow rate to the system decontamination. In this paper, the flow characteristics in Kori-1 NPP reactor coolant according to RHR pump operation were evaluated. The evaluation results showed that system decontamination using an RHR pump was not effective at decontamination due first to impurities deposited in piping and equipment, and second to the extreme flow unbalance in the RCS caused deposition of impurities.