• The Transactions of The Korean Institute of Electrical Engineers
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• v.65 no.5
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• pp.836-842
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• 2016

When an accident occurs in the nuclear power plant, the faulted information might mislead to the high possibility of aggravating the accident. At the Fukushima accident, the operators misunderstood that there was no core exposure despite in the processing of core damage, because the instrument information of the reactor water level was provided to the operators optimistically other than the actual situation. Thus, this misunderstanding actually caused to much confusions on the rapid countermeasure on the accident, and then resulted in multiplying the accident propagation. It is necessary to be equipped with the function that informs operators the status of instrument integrity in real time. If plant operators verify that the instruments are working properly during accident conditions, they are able to make a decision more safely. In this study, we have performed various tests for the fault detection sensitivity of an data-driven empirical model to review the usability of the model in the accident conditions. The test was performed by using simulation data from the compact nuclear simulator that is numerically simulated to PWR type nuclear power plant. As a result of the test, the proposed model has shown good performance for detecting the specified instrument faults during normal plant conditions. Although the instrument fault detection sensitivity during plant accident conditions is lower than that during normal condition, the data-drive empirical model can be detected an instrument fault during early stage of plant accidents.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.2
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• pp.235-243
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• 2019

Radioactive materials are released into the air and deposited on the surface soil after a nuclear accident. Radionuclides deposited in soil are transported by precipitation to nearby environments and contaminate the surface water system. Basic data on surface watershed and soil erosion models have been collected and analyzed to evaluate the behavior of radionuclides deposited on surface soil after a nuclear accident. Data acquisition and analysis in aquatic environment were performed to investigate the physical characteristics and variation of biota in rivers and lakes of the Nakdong river area near the Wolsong nuclear power plant. For these purposes, a digital map, and hydrological, water quality and biota data were gathered and a systematic database (DB) was constructed in connection with them. Constructed aquatic DB will be supplied and used in surface watershed and soil erosion models for investigation of long-term movement of radionuclides in adsorptive form in surface soil. Finally, basic data and established models will be utilized for general radiological impact assessment in aquatic environment.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.364-370
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• 1995

The One-dimensional fin model is used to analyze the angular wall temperature variation of long horizontal lines, where stratification could result in top-to-bottom differences in wall temperatures. The top and bottom sections are treated separately and coupled by boundary conditions. The thermal stratification analysis is focused on the effects of the heat transfer rates at the pipe surface. The results show that the heat transfer rate at the pipe surface is the controlling parameter which reduce significantly the temperature difference in pipe circumferential direction. The one-dimensional fin modelling analysis results are verified by comparison with the operating PWR test data. The circumferential temperatures of pipe calculated by one-dimensional fin modelling agree well with the PWR plant test data.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.290-295
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• 1996

The purpose of this study is to establish the pressure/temperature curves of Reactor Coolant System for brittle fracture prevention. The pressure/temperature curve is the basis to select RC Pump and limits to operate the plant. Based on the plant operation experience, this curve should be re-generated periodically in order to ensure the structural integrity using data from the test of reactor vessel surveilance materials to compensate for the irradiation effects. This study provides the procedure of pressure/temperature curve generation in term of brittle fracture prevention of reactor vessel. Using the UCN 3&4 data, the sample pressure/temperature curve was generated, and it was compared with those of YGN 3&4 based on the stress and $RT_{NDT}$value.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.453-457
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• 1998

The sorption of UO$_2$$^{2+}$showed ionic strength independece for goethite and dependence for kaolinite. In the presence of carbonate, the sorption decreased in the weakly alkaline pH range becase of the formation of aqueous U(VI)-carbonate complexes. The sorption of UO$_2$$^{2+}$onto goethite and kaolinite under various experimental conditions was successfully interpreted using a surface complexation modeling, named triple layer model (TLM). The best fit to the experimental data was obtained by the FITEQL program, and then evaluated with available spectroscopic data. The results showed the versatility of surface complexation modeling over empirical one to predict UO$_2$$^{2+}$ sorption behavior.avior.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.745-750
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• 1996

An analysis of a loss of residual heat removal system (RHRS) event during midloop operation after reactor shutdown was performed using the RELAP5/MOD3 thermal-hydraulic computer code. The experimental data of a 5% cold leg break test conducted at the ROSA-IV Large Scale Test Facility (LSTF) to simulate a main coolant pump shaft seal removal event during midloop operation of a Westinghouse-type PWR were used in the analysis. The predicted core boiling time and the peak primary system pressure showed good agreements with the measured data. Some differences between the calculational results and the experimental results were, however, found in areas of the timing of loop seal clearing and the temperature distribution in a pressurizer. Other calculational problems identified were discussed as well.

• Nuclear Engineering and Technology
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• v.35 no.1
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• pp.25-34
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• 2003

The research on the development of the fault monitoring system for the thermal reduction reactor has been performed preliminarily in order to support the successful operation of the thermal reduction reactor. The final task of the development of the fault monitoring system is to assure the integrity of the thermal$_3$ reduction reactor by the acoustic emission (AE) method. The objectives of this paper are to identify and characterize the fault-induced signals for the discrimination of the various AE signals acquired during the reactor operation. The AE data acquisition and analysis system was constructed and applied to the fault monitoring of the small- scale reduction reactor, Through the series of experiments, the various signals such as background noise, operating signals, and fault-induced signals were measured and their characteristics were identified, which will be used in the signal discrimination for further application to full-scale thermal reduction reactor.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.532-537
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• 1998

Friction factor and two-phase flow frictional multiplier for a CANFLEX bundle are newly developed and presented in this paper. CANFLEX as a 43-element fuel bundle has been developed jointly by AECL/KAERI to provide greater operational flexibility for CANDU reactor operators and designers. Friction factor and two-phase flow frictional multiplier have been developed by using the experimental data of pressure drops obtained from two series of Freon-l34a (R-134a) CHF tests with a string of simulated CANFLEX bundles in a single phase and a two-phase flow conditions. The friction factor for a CANFLRX bundle is found to be about 20 % higher than that of Blasius for a smooth circular pipe. The pressure drop predicted by using the new correlations of friction factor and two-phase frictional multiplier are well agreed with the experimental pressure drop data of CANFLEX bundle within ${\pm}\;5\;%$ error.

• Nuclear Engineering and Technology
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• v.32 no.3
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• pp.227-234
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• 2000

It is known that loose parts in the reactor coolant system (RCS) cause serious damage to the systems. This paper is concerned with estimating the mass of a loose part in the steam generator of a nuclear power plant. We developed the mass estimation algorithm based on the Hertz theory in order to estimate the mass of the loose parts and applied the algorithm to the impact test data of YGN3. The mass estimation values were compared with real values in order to verify the algorithm. The result showed that the average error of the mass estimation value is less than 27%.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.522-527
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• 1997

A total of 386 critical flow data points from 19 runs of 27 runs in the Marviken Test were selected and compared with the predictions by the correlations based on the critical flow scaling parameters. The results show that the critical mass flux in the very large diameter pipe can be also characterized by two scaling parameters such as discharge coefficient and dimensionless subcooling( $C_{d, ref}$ and $\Delta$ $T^{*}$$_{sub}$). The agreement between the measured data and the predictions are excellent.t.ons are excellent.t.