• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3563-3579
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• 2021

Measured decay heat data of light water reactor (LWR) spent nuclear fuel (SNF) assemblies are adopted to train machine learning (ML) models. The measured data is available for fuel assemblies irradiated in commercial reactors operated in the United States and Sweden. The data comes from calorimetric measurements of discharged pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies. 91 and 171 measurements of PWR and BWR assembly decay heat data are used, respectively. Due to the small size of the measurement dataset, we propose: (i) to use the method of multiple runs (ii) to generate and use synthetic data, as large dataset which has similar statistical characteristics as the original dataset. Three ML models are developed based on Gaussian process (GP), support vector machines (SVM) and neural networks (NN), with four inputs including the fuel assembly averaged enrichment, assembly averaged burnup, initial heavy metal mass, and cooling time after discharge. The outcomes of this work are (i) development of ML models which predict LWR fuel assembly decay heat from the four inputs (ii) generation and application of synthetic data which improves the performance of the ML models (iii) uncertainty analysis of the ML models and their predictions.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1532-1539
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• 2019

This paper analyzes a new method for reducing boiling water reactor fuel temperature during a Loss of Coolant Accident (LOCA). The method uses a device called Reverse Flow Restriction Device (RFRD) at the inlet of fuel bundles in the core to prevent coolant loss from the bundle inlet due to the reverse flow after a large break in the recirculation loop. The device allows for flow in the forward direction which occurs during normal operation, while after the break, the RFRD device changes its status to prevent reverse flow. In this paper, a detailed simulation of LOCA has been carried out using the U.S. NRC's TRACE code to investigate the effect of RFRD on the flow rate as well as peak clad temperature of BWR fuel bundles during three different LOCA scenarios: small break LOCA (25% LOCA), large break LOCA (100% LOCA), and double-ended guillotine break (200% LOCA). The results demonstrated that the device could substantially block flow reversal in fuel bundles during LOCA, allowing for coolant to remain in the core during the coolant blowdown phase. The device can retain additional cooling water after activating the emergency systems, which maintains the peak clad temperature at lower levels. Moreover, the RFRD achieved the reflood phase (when the saturation temperature of the clad is restored) earlier than without the RFRD.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.793-803
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• 2021

The small PWR has been paid more and more attention due to its diversity of application and flexibility in the site selection. However, the large core power density, the small containment space and the rapid accident progress characteristics make it difficult to control the containment pressure like the traditional PWR during the LOCA. The pressure suppression system has been used by the BWR since the early design, which is a suitable technique that can be applied to the small PWR. Since the configuration and operating conditions are different from the BWR, the pressure suppression system should be redesigned for the small PWR. Conducting the experiments on the scale down test facility is a good choice to reproduce the prototypical phenomena in the test facility, which is both economical and reasonable. A systematic scaling method referring to the H2TS method was proposed to determine the geometrical and thermohydraulic parameters of the pressure suppression containment response test facility for the small PWR conceptual design. The containment and the pressure suppression system related thermohydraulic phenomena were analyzed with top-down and bottom-up scaling methods. A set of the scaling criteria were obtained, through which the main parameters of the test facility can be determined.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1296-1309
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• 2024

The aim of this work is to explore the effect of the double subdiffusion on the stability in BWRs. A BWR novel reduced order model with double subdiffusion effects: reduced order fractional model (DS-F-ROM) to describe the neutron and heat transfer processes was proposed for this study. The double subdiffusion was developed with a fractional-order two-equation model, and with different fractional-orders and relaxation times. The stability analysis was carried out using the root-locus method and change from the s to the W domain and were confirmed using the time-domain evolution of neutron flux for a unit step change in reactivity. The results obtained using the reduced fractional-order model are presented for different anomalous diffusion coefficient values. Results are compared with normal diffusion and P1 equations, which are obtained straightforwardly with DS-ROM when relaxation time tends to zero, and when the anomalous diffusion coefficient tends to one, respectively.

• 제어로봇시스템학회:학술대회논문집
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• 1988.10b
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• pp.804-809
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• 1988

PRNM(Power Range Neutron Monitoring) of BWR (Boiling Water Reactor) is a system that processes signals from about two hundred LPRM (Local Power Range Monitor) sensors in the nuclear reactor and this system monitors the neutron flux level during the plant operating region. Development has been made by employing a special technique for multiplexing neutron sensor signals and the recent advanced microelectronics technology. It is applicable to the total plant digital control system for a nuclear power plant.

• Proceedings of the KWS Conference
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• 2004.05a
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• pp.308-310
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• 2004

1980년대 초 미국의 비등형 경수로(Boiling Water Reactor : BWR) 원자력발전소 배관계통의 입계 응력 부식 균열(Inter-Granular Stress Corrosion Crack) 검사 결과 및 미국 EPRI(Electric Power Research Institute)에서 실시한 round robin test 결과에서 기존 초음파 검사 방법의 실효성에 많은 문제점이 제기 되었다. (중략)

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05b
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• pp.1087-1092
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• 1995

MOX 연료는 PWR, BWR, FBR 및 ATR 등과 같은 다양한 노형에서 이용이 가능하다. 이와 같은 MOX 연료는 이미 일부 국가에서 이들 노형에 이용하고 있으며, FBR의 상용화 시기가 다소 지연될 것으로 예상되어 이 분야에 대찬 연구는 더욱 활발히 진행될 전망이다. 여기에서는 이와 같은 현실을 감안하여 MOX 연료를 이용하는 MOX 연료주기와 우라늄 연료만을 이용하는 우라늄 연료주기에 대한 핵연료주기비를 비교·분석하였고, 그 결과 MOX 연료주기와 우라늄 연료주기의 경제성은 거의 대둥한 것으로 평가되었다. 또한 우라늄 가격, 사용후핵연료 처분시기, 처분 비용/재처리비용 등의 주요 변수에 대한 민감도분석을 수행한 결과, MOX 연료주기가 우라늄연료주기에 대해 상당한 경쟁력을 가지고 있는 것으로 분석되었다.

• Journal of the Korean Institute of Gas
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• v.1 no.1
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• pp.41-48
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• 1997

The equation of state is widely utilized as a simple model for the prediction of gas properties. There are several equations of state and they often make diverse and hard to believe output of gas properties. In this study, We show a reliability of equation of state for nitrogen, oxygen and argon in pressure range from 1 bar to 30 bar and temperature range from liquefaction to room temperature. We use three equations of state such as Soave-Redlich-Kwong, Peng-Robinson and BWR-LS' equation of state which provided in the Aspen plus. The results were compared with literatures and virial equation. Finally, We report the differences of process calculation of distillation column and expansion turbine in cryogenic air separation plant with change of equation of state.

• 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.

• Journal of the Korean Society for Nondestructive Testing
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• v.13 no.2
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• pp.48-60
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• 1993

1974년에 유럽에서 시작된 PISC(Program for the Inspection of Steel Components ; PISC-I, II, III) 수행결과나 1980년대초 실시된 미국의 BWR 원전 배관계통의 입계응력부식균열(IGSCC ; Intergranular Stress Corrosion Cracks)검사결과에서 나타난 바와 같이 기존의 규격 요건과 절차에 따른 원자력발전소 기기에 대한 초음파탐상검사는 그 실효성에 많은 문제점이 제기되었다. 따라서, 원전기기의 건전성 및 초음파탐상검사 결과의 신뢰도를 보증하기 위한 각종 연구가 진행되고 여러 방안이 모색되어 왔다. 그 결과, 원전 가동중검사 규격에 "초음파탐상검사자 자격인정 요건"과 "초음파탐상검사 시스템(검사자, 장비 및 절차서)에 대한 기량검증 요건"이 새로이 부가되었다. 본고에서는 초음파탐상검사 결과의 신뢰도 확보에 있어 필수불가결한 요건인 원전기기 초음파 탐상검사 시스템에 대한 기량검증 요건을 자세히 기술한다.