• Nuclear Engineering and Technology
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• 제52권5호
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• pp.878-888
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• 2020

This paper presents the validation of UNIST in-house Monte Carlo code MCS used for the high-fidelity simulation of commercial pressurized water reactors (PWRs). Its focus is on the accurate, spatially detailed neutronic analyses of startup physics tests for the initial core of the Watts Bar Nuclear 1 reactor, which is a vital step in evaluating core phenomena in an operating nuclear power reactor. The MCS solutions for the Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) core physics benchmark progression problems 1 to 5 were verified with KENO-VI and Serpent 2 solutions for geometries ranging from a single-pin cell to a full core. MCS was also validated by comparing with results of reactor zero-power physics tests in a full-core simulation. MCS exhibits an excellent consistency against the measured data with a bias of ±3 pcm at the initial criticality whole-core problem. Furthermore, MCS solutions for rod worth are consistent with measured data, and reasonable agreement is obtained for the isothermal temperature coefficient and soluble boron worth. This favorable comparison with measured parameters exhibited by MCS continues to broaden its validation basis. These results provide confidence in MCS's capability in high-fidelity calculations for practical PWR cores.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4551-4559
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• 2022

The aim of this research is to analyze the characteristics of a core melt discharged from the reactor vessel and the spreading behavior the core melt in the reactor cavity of the SMART. First, a severe accident sequence under conservative conditions is simulated by the MELCOR code to obtain the conditions for an analysis of the spreading behavior and coolability of the ex-vessel melt. Second, the spreading behavior and coolability of the ex-vessel melt are analyzed by the MELTSPREAD code. The level, temperature, and pressure of the water in the cavity as well as the temperature, mass, composition, and discharge velocity of the melt were utilized to construct the ex-vessel analysis. The melt spread only to part of the cavity, and that the height of the corium in a static state was less than 25 cm. The characteristics of a small modular reactor on the spreading behavior and coolability of melt were analyzed. In the SMART, the amount of melt discharged into the cavity is relatively small and the area of the cavity is sufficiently large when compared to a high-power pressurized water reactor. It was found that the coolability of an ex-vessel core melt can be sufficiently secured.

• 한국압력기기공학회 논문집
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• 제10권1호
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• pp.44-50
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• 2014

There were many flaw issues of reactor vessel head penetration in USA fleets. USNRC issued 10CFR50.55a to implement reactor vessel head penetration ultrasonic examination performance demonstration(PD) in US for enhancement of inspection reliability. After September 2009, all US utilities inspected their RVHP with PD qualified system. Korea Hydro and Nuclear Power Company(KHNP) have developed reactor vessel head penetration performance demonstration system for ultrasonic test to apply for pressurized light-water reactor power plants in accordance with 10CFR50.55a since September 2011. RVHP configuration surveying and analysis, code requirement analysis, and performance demonstration specimen design were performed up to this day. Fingerprinting of manufactured specimen, development of test data management program, development of operation procedure, input of flawed data, and development of final report will be performed for the next step. This paper describes the development status of the performance demonstration system for reactor vessel head penetration ultrasonic examination in Korea.

• Journal of Advanced Marine Engineering and Technology
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• 제28권5호
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• pp.827-836
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• 2004

The design parameter of the heat loss for the pressurized water reactor has been studied. The heat loss from the reactor vessel through the air gap. insulation are analysed by using the computational fluid dynamics code, FLUENT. Parametric study has been performed on the air gap width between the reactor vessel wall and the inner surface of the insulation, and on the insulation thickness. Also evaluated is the performance degradation due to the chimney effect due to gaps left between the panels during the installation of the insulation system. From the analysis results, the optimal with of air gap and insulation thickness and the value of heat loss are obtained The results show how the heat loss varies with the air gap width and insulation thickness. The temperature and the velocity distributions are also presented. From the results of the evaluation. the optimal air gap width and the optimal insulation thickness are obtained. As the difference between the predicted heat loss and measured heat loss from the reactor vessel is construed Primarily as losses due to chimney effect. the contribution of the chimney effect to the total heat loss is quantitatively indicated.

• Nuclear Engineering and Technology
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• 제52권7호
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• pp.1429-1435
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• 2020

Applying an accurate parametric prediction model to identify abnormal or false pressurizer water levels (PWLs) is critical to the safe operation of marine pressurized water reactors (PWRs). Recently, deep-learning-based models have proved to be a powerful feature extractor to perform high-accuracy prediction. However, the effectiveness of models still suffers from two issues in PWL prediction: the correlations shifting over time between PWL and other feature parameters, and the example imbalance between fluctuation examples (minority) and stable examples (majority). To address these problems, we propose a cost-sensitive mechanism to facilitate the model to learn the feature representation of later examples and fluctuation examples. By weighting the standard mean square error loss with a cost-sensitive factor, we develop a Cost-Sensitive Long Short-Term Memory (CSLSTM) model to predict the PWL of PWRs. The overall performance of the CSLSTM is assessed by a variety of evaluation metrics with the experimental data collected from a marine PWR simulator. The comparisons with the Long Short-Term Memory (LSTM) model and the Support Vector Regression (SVR) model demonstrate the effectiveness of the CSLSTM.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집D
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• pp.343-347
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• 2001

Advanced Pressurized Reactor 1400(APR-1400), which is a standard evolutionary advanced light water reactor(ALWR), has been developed from 1992 as one of long-term Government Project(G-7). The APR-1400 is designed to operate at the rated output of 4000MWt to produce an electric power output of around 1450MWe. The balance of plant (BOP) for the secondary system consists of main steam, feedwater, condensate, turbine generator and auxiliary system. In this paper, we describe the major design features of secondary component, balance of plant configuration, and then the turbine cycle thermal performance evaluation using PEPSE code.

• Nuclear Engineering and Technology
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• 제19권4호
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• pp.279-291
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• 1987

현재 원자력발전소의 확률론적 위해도 평가에 사용되는 사상 수목이나 고장수목 기법은 부품이나 계통의 이원적상태와 정적 묘사에 근거하고 있다 이 기법이 대부분의 안전해석에는 적합하지만, 요사이 점차 중요관심사가 되고 있는 발전소의 이용률 측정이나 기술 사양서 평가 같은 문제를 정확하게 다루기 위해서는 마코프 신뢰도 분석과 같은 보다 진보된 기법이 필요하다. 이 논문은 가압경수로의 원자로 보호계통을 위한 마코프 신뢰도 모델을 기술하고 기술사양서의 두 검사 절차를 분석한 결과를 제시한다.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 1993년도 봄 학술발표회논문집
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• pp.159-164
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• 1993

This paper presents a computer design program for In-Core Instrumentation(ICI) guide tube routing and locations on support system, and checking the interference between ICI guide tubes in the reactor coolant system of typical Pressurized Water Reactor. The program, ICITRIC, has been written in FORTRAN language which is available under UNIX environment. Results of this program are compared with those of the commercial code, PATRAN, and both results are almost same Also the results may provide input data for ICI system static and dynamic analysis performed by the commercial code, SUPER PIPE. This program can simulate ICI guide tube routing and locations on support system, and checking the interference between ICI guide tubes. Through a process of iteration, the designer can apply initial conditions, and modify the routing until satisfied with the overall system performance.

• Nuclear Engineering and Technology
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• 제44권5호
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• pp.501-506
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• 2012

Many innovative design features are employed in the reactor vessel internals of SMART, a small integral-type pressurized water reactor, one of which is the flow skirt, which uniformly distributes flow and horizontally restrains the lower part of the core support barrel. This new design requires a comprehensive investigation of vibration characteristics. Therefore, in this study, modal characteristics of flow skirts are investigated with finite element analysis. Specifically, we investigate how the presence of holes, the presence of three rings attached to the flow skirt, and the thickness of the lowest shell effect vibration characteristics. In addition, the fluid effect is addressed, since the flow skirt is submerged in the fluid.

• Nuclear Engineering and Technology
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• 제36권2호
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• pp.175-183
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• 2004

For the analysis of a dry process fuel cycle, new modules were implemented into the fuel cycle analysis code DYMOND, which was developed by the Argonne National Laboratory. The modifications were made to the energy demand prediction model, a Canada deuterium uranium (CANDU) reactor, direct use of spent pressurized water reactor (PWR) fuel in CANDU reactors (DUPIC) fuel cycle model, the fuel cycle calculation module, and the input/output modules. The performance of the modified DYMOND code was assessed for the postulated once-through fuel cycle models including both the PWR and CANDU reactor. This paper presents modifications of the DYMOND code and the results of sample calculations for the PWR once-though and DUPIC fuel cycles.