• Nuclear Engineering and Technology
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• v.18 no.3
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• pp.175-182
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• 1986

The KNUI (Korea Nuclear Unit 1) loss of offsite power transient as a design-base accident has been simulated using the RELAP5/MOD1/NSC computer code. The analysis is carried out using the best-estimate methodology, but the sequence and its assumptions are based on the evaluation methodology th at emphasizes conservatism. Important thermal-hydraulic parameters such as average temperature, steam generator level and pressurizer water volume are compared with the results in the KNU1 Final Safety Analysis Report (FSAR). The present analysis gives much lower RCS average temperature and pressurizer water volume, and much higher S/G water volume at the turnaround point, which may be considered to be additional improved safety margins. This is expected since the present analysis deals with the best-estimate thermal-hydraulic models as well as the initial conditions on a best-estimate basis. These additional safety margins may contribute to further validate the safety of the KNU1 in this type of accidents(Decrease in Heat Removal by the Secondary System).

• Nuclear Engineering and Technology
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• v.24 no.1
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• pp.30-39
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• 1992

In this study, an integrated boration and dilution (INBAD) model is proposed to predict the required makeup flowrate for RCS boron concentration change and to analyze the boron concentration behavior at each subsystem within the RCS including CVCS during boration and dilution operation. The INBAD model is constructed by integrating an existing neutronic code and a boration and dilution model. The boration and dilution model has been developed for our specific purpose using the one-cell model and multi-cell model. In addition, in order to assess the boron concentration behavior more realistically, two important features such as variable pressurizer heater output and optional makeup mode (either direct or indirect injection) are implemented in this model. In order to demonstrate the usefulness of this model, the boron concentration behavior analysis at each subsystem were performed for both direct and indirect injection mode using YGN 3 and 4 design data. Also, the effect of pressurizer heater output on the primary loop boron concentration was investigated. The results showed that the boron concentration changes can be predicted accurately at each subsystem during boration and dilution operation.

• Nuclear Engineering and Technology
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• v.22 no.1
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• pp.12-19
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• 1990

The Loss of Offsite Power Transient at 77.5% power which occurred on June 9, 1981 at the Kori Unit 1 PWR (Pressurized Water Reactor) is simulated using the RELAP5/MOD2 system thermal-hydraulics computer code. Major thermal-hydraulic parameters are compared with the available plant data. The comparison of the analysis results with the plant data demonstrates that the RELAP5/MOD2 code has the capability to simulate the thermal-hydraulic behaviour of PWRs under accident conditions of this type with accuracy, except the pressurizer pressure and level. The pressurizer pressure increase is sensitive to the in surge now it is believed that the interracial heat transfer in a horizontal stratified flow regime may be estimated low and the compression effect due to insurge flow may be high. In the nodalization sensitivity study it is found that S/G noding with junctions between bypass plenum and steam dome is preferred to simulate the S/G water level decreasing and avoid the spurious level peak at trubine trip.

• Nuclear Engineering and Technology
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• v.33 no.5
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• pp.483-497
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• 2001

In this work, a neuro-fuzzy inference system combined with the wavelet denoising, PCA (principal component analysis) and SPRT (sequential probability ratio test) methods is developed to detect the relevant sensor failure using other sensor signals. The wavelet denoising technique is applied to remove noise components in input signals into the neuro-fuzzy system The PCA is used to reduce the dimension of an input space without losing a significant amount of information. The PCA makes easy the selection of the input signals into the neuro-fuzzy system. Also, a lower dimensional input space usually reduces the time necessary to train a neuro-fuzzy system. The parameters of the neuro-fuzzy inference system which estimates the relevant sensor signal are optimized by a genetic algorithm and a least-squares algorithm. The residuals between the estimated signals and the measured signals are used to detect whether the sensors are failed or not. The SPRT is used in this failure detection algorithm. The proposed sensor-monitoring algorithm was verified through applications to the pressurizer water level and the hot-leg flowrate sensors in pressurized water reactors.

• Nuclear Engineering and Technology
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• v.32 no.5
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• pp.485-494
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• 2000

In this work, a fuzzy neural network is used to estimate the relevant sensor signal using other sensor signals. Noise components in input signals into the fuzzy neural network are removed through the wavelet denoising technique . Principal component analysis (PCA) is used to reduce the dimension of an input space without losing a significant amount of information. A lower dimensional input space will also usually reduce the time necessary to train a fuzzy-neural network. Also, the principal component analysis makes easy the selection of the input signals into the fuzzy neural network. The fuzzy neural network parameters are optimized by two learning methods. A genetic algorithm is used to optimize the antecedent parameters of the fuzzy neural network and a least-squares algorithm is used to solve the consequent parameters. The proposed algorithm was verified through the application to the pressurizer water level and the hot-leg flowrate measurements in pressurized water reactors.

• 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.

• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.551-556
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• 1997

A present study addresses a loss-of-RHR event as an initiating event under specific low power or shutdown conditions. Two typical plant configurations, cold leg opening case with water-filled steam generators and pressurizer opening case with emptied steam generators, were evaluated using the RELAP5/ MOD3.2 code. The calculation was compared with the experiment conducted at ROSA-IV/LSTF in Japan. As a result, the code was capable of simulating the system transient behavior following the event. Especially, thermal hydraulic transport processes including non-condensable gas behavior were reasonably predicted with an appropriate time step and CPU time. However, there were some code deficiencies such as too large system mass errors and severe flow oscillations in core region.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.324-329
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• 2004

In this study, an application of crack morphology variables in the Leak-Before-Break(LBB) evaluation for nuclear piping systems is investigated, including influences on the leakage crack size and crack instability loads. The crack surface roughness and the number of flow turns as a function of the crack opening displacement are applied to LBB evaluations for KSNP pressurizer surge line, for which fatigue and stress corrosion cracking are considered as failure mechanisms. As a result, there would be a significant impact on safety margins to acceptance criteria for the surge line if crack morphology variables are applied additionally to the current regulatory guide without re-analyses for justification of safety factors being applied on the leakage crack size and piping loads for evaluations.

• Nuclear Engineering and Technology
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• v.40 no.3
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• pp.239-250
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• 2008

If there is a water flow with a range of temperature inside a pipe, the wanner water tends to float on top of the cooler water because it is lighter, resulting in the upper portion of the pipe being hotter than the lower portion. Under these conditions, such thermal stratification can play an important role in the aging of nuclear power plant piping because of the stress caused by the temperature difference and the cyclic temperature changes. This stress can limit the lifetime of the piping, even leading to penetrating cracks. Investigated in this study is the effect of thermal stratification on the structural integrity of the pressurizer surge line, which is reported to be one of the pipes most severely affected. Finite element models of the surge line are developed using several element types available in a general purpose structural analysis program and stress analyses are performed to determine the response characteristics for the various types of top-to-bottom temperature differentials due to thermal stratification. Fatigue analyses are also performed and an allowable environmental correction factor is suggested.