• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4628-4636
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• 2023

Cables are indispensable in nuclear power plants for transmitting data measured by various types of detectors, such as self-powered neutron detectors (SPNDs). These cables will generate disturbing signals that must be accurately distinguished and eliminated. Given that the cable current is not very significant, previous research has focused on SPND, with little attention paid to cable evaluation and validation. This paper specifically focuses on the quantitative analysis of cables and proposes a theoretical model to predict cable noise. In this model, the reaction characteristics between irradiated neutrons and cables were discussed thoroughly. Based on the Monte Carlo method, a comprehensive simulation approach of neutron sensitivity was introduced and long-term irradiation experiments in a heavy water reactor (HWR) were designed to verify this model. The theoretical results of this method agree quite well with the experimental measurements, proving that the model is reliable and exhibits excellent accuracy. The experimental data also show that the cable current accounts for approximately 0.2% of the total current at the initial moment, but as the detector gradually depletes, it will contribute more than 2%, making it a non-negligible proportion of the total signal current.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2576-2583
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• 2024

In this paper, the similarity criterion and dimensionless conversion method combined with the elasticity condition and Hooke's law are used to derive the functional relationship of the maximum effective value of the vibration velocity between the prototype pump and the model pump. The seawater circulation pump of a nuclear power plant is used as the prototype pump, and the model pump is obtained by performance conversion and choosing the appropriate scale, and the vibration state of the model pump under different flow rates is measured and analyzed. The vibration data of the model pump through the function relationship to find out the vibration parameters of the prototype model pump, and compare with the vibration data of the seawater circulation pump in reality. It can be seen that with the increase of flow rate, the maximum effective value of the vibration velocity of both model and prototype decreases and then increases, and the relative error is small, the maximum value is 7.7757%. Therefore, it can be considered that the functional relationship of model pump converted to prototype pump derived in this paper can be used to analyze the vibration of the actual seawater circulation pump of coastal nuclear power plant.

• Proceedings of the KIEE Conference
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• 1999.07b
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• pp.1013-1016
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• 1999

In this paper, a real-time communication method using a PICNET-NP(Plant Instrumentation and Control Network for Nuclear Power plant) is proposed with an analysis of the control network requirements of DCS (Distributed Control System) in nuclear power plants. The method satisfies deadline in case of worst data traffics by considering aperiodic and periodic real-time data and others. In addition, the method was used to analyze the data characteristics of the DCS in existing nuclear power plant. The result shows that use of this method meets the response time requirement(100ms)

• Nuclear Engineering and Technology
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• v.43 no.2
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• pp.99-104
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• 2011

A Structure Health Monitoring (SHM) system needs a real-time remote data acquisition system to monitor the status of a structure from anywhere via Internet access. In this paper, we present a data acquisition system that monitors up to 40 Fiber Bragg Grating Sensors remotely in real-time. Using a TCP/IP protocol, users can access information gathered by the sensors from anywhere. An experiment in laboratory conditions has been done to prove the feasibility of our proposed system, which is built in special-purpose monitoring system.

• International conference on construction engineering and project management
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• 2017.10a
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• pp.133-137
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• 2017

Nuclear power plants in Korea are usually built based on a duplicated model; so the project cost data of the preceding unit can be used as reference when estimating the project cost for the succeeding unit. However, since the contracting method is oriented towards the price, empirical factors such as making top-down estimations using the reverse calculation method based on the completion cost of the preceding unit is dominant. In order to develop a project cost database to resolve such problems, the detailed cost boundary of the project cost data must be categorized by project and by system. This study proposes a method to connect the code of account with the base quantities and the IAEA account, and proposes a database structure for the development of a project cost estimation system. The estimation system developed in the future is expected to utilize the proposed project cost data structure.

• Nuclear Engineering and Technology
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• v.39 no.3
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• pp.215-220
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• 2007

The existing flooding Probabilistic Safety Analysis(PSA) was updated to reflect the Korean plant specific operating experience data into the flooding frequency to improve the PSA quality. Both the Nuclear Power Experience(NPE) database and the Korea Nuclear Pipe Failure Database(NuPIPE) databases were used in this study, and from these databases, only the Pressurized Water Reactor(PWR) data were used for the flooding frequencies of the flooding areas in the primary auxiliary building. With these databases and a Bayesian method, the flooding frequencies for the flooding areas were estimated. Subsequently, the Core Damage Frequency(CDF) for the flooding PSA of the Ulchin(UCN) unit 3 and 4 plants based on the Korean Standard Nuclear Power Plant(KSNP) internal full-power PSA model was recalculated. The evaluation results showed that sixteen flooding events are potentially significant according to the screening criterion, while there were two flooding events exceeding the screening criterion of the existing UCN 3 and 4 flooding PSA. The result was compared with two kinds of cases: (1) the flooding frequency and CDF from the method of the existing flooding PSA with the PWR and Boiled Water Reactor(BWR) data of the NPE database and the Maximum Likelihood Estimate(MLE) method and (2) the flooding frequency and CDF with the NPE database(PWR and BWR data), NuPIPE database, and a Bayesian method. From the comparison, a difference in CDF results was revealed more clearly between the CDF from this study and case (2) than between case (1) and case (2). That is, the number of flooding events exceeding the screen criterion further increased when only the PWR data were used for the primary auxiliary building than when the Korean specific data were used.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2610-2615
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• 2021

The prediction of irradiation-induced transition temperature shift for RPV steels is an important method for long term operation of nuclear power plant. Based on the irradiation embrittlement data, an irradiation-induced transition temperature shift prediction model is developed with machine learning method XGBoost. Then the residual, standard deviation and predicted value vs. measured value analysis are conducted to analyze the accuracy of this model. At last, Cu content threshold and saturation values analysis, temperature dependence, Ni/Cu dependence and flux effect are given to verify the reliability. Those results show that the prediction model developed with XGBoost has high accuracy for predicting the irradiation embrittlement trend of RPV steel. The prediction results are consistent with the current understanding of RPV embrittlement mechanism.

• Proceedings of the Korea Concrete Institute Conference
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• 1998.10b
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• pp.637-642
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• 1998

This study was performed to develop effective management system of concrete structures in Nuclear Power Plants. This D/B system includes three kinds of data : 1)visual inspection data(cracking, spalling, etc.) 2) durability data carbonation, chloride attack, etc. 3) in-service inspection data(prestressing force. material properties, etc. ) By using the life management D/B System, the field engineers can easily acquire the information about the various inspection data. repair and accidental histories of structures. This system, will contribute to the efficient life management of concrete structures.

• Nuclear Engineering and Technology
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• v.15 no.1
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• pp.41-49
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• 1983

A computer code, BERD (Bayesian Estimation of Reliability Data), has been developed and tested in order to update the data for the reliability analysis of safety related systems in a specific nuclear power plant. The code has been used to derive the plant-specific data for reliability analysis of the auxiliary feedwater system of a pressurized water reactor. The prior information for components selected was taken from the U.S. Reactor Safety Study, WASH-1400, and the operating experiences from published licensee event reports. The results show that the updated data are well fitted to log-normal distribution curves and the error factors are reduced significantly.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1616-1629
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• 2023

To quantify the conservatism of existing ASME strain-based evaluation methods for seismic loading, this paper presents very low cycle fatigue test data of elbows under various cyclic loading conditions and comparison of evaluation results with experimental failure cycles. For strain-based evaluation methods, the method presented in ASME BPVC CC N-900 and Sec. VIII are used. Predicted failure cycles are compared with experimental failure cycle to quantify the conservatism of evaluation methods. All methods give very conservative failure cycles. The CC N-900 method is the most conservative and prediction results are only ~0.5% of experimental data. For Sec. VIII method, the use of the option using code tensile properties gives ~3% of experimental data, and the use of the material-specific reduction of area can reduce conservatism but still gives ~15% of experimental data.