• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2878-2887
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• 2021

As the amount of fission product released from ISLOCA was overestimated because of conservative assumptions in the past, several studies have been recently conducted to evaluate the actual release amount. Among several pathways for the ISLOCA, most studies were focused on the pathway with the highest possibility. However, different ISLOCA pathways may have different fission product release characteristics. In this study, fission product behavior was analyzed for various pathways at the Westinghouse two-loop plant using MELCOR. Four pathways are considered: the pipes from a cold leg, from a downcomer, from a hot leg to the outlet of RHR heat exchanger, and the pipe from the hot leg to the inlet of RHR pump (Pathway 1-4). According to the analysis results, cladding fails at around 2.5 h in Pathways 1 and 2, and on the other hand, about 3.3 h in Pathways 3 and 4 because the ISLOCA pathways affect the safety injection flow path. While the release amount of cesium and iodine ranges between 20 and 26% in Pathways 1 to 3, Pathway 4 allows only 5% to the environment because the break location is submerged. Also, as more than 90% of cesium released to the environment passes through the personnel door, reinforcing the pressure capacity of the doors would be a significant factor in the accident management of the ISLOCA.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.356-368
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• 2019

This paper presents the verification and validation (V&V) of the STREAM/RAST-K 2.0 code system for a pressurized water reactor (PWR) analysis. A lattice physics code STREAM and a nodal diffusion code RAST-K 2.0 have been developed by a computational reactor physics and experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) for an accurate two-step PWR analysis. The calculation modules of each code were already verified against various benchmark problems, whereas this paper focuses on the V&V of linked code system. Three PWR type reactor cores, OPR-1000, three-loop Westinghouse reactor core, and APR-1400, are selected as V&V target plants. This code system, for verification, is compared against the conventional code systems used for the calculations in nuclear design reports (NDRs) and validated against measured plant data. Compared parameters are as follows: critical boron concentration (CBC), axial shape index (ASI), assembly-wise power distribution, burnup distribution and peaking factors. STREAM/RAST-K 2.0 shows the RMS error of critical boron concentration within 20 ppm, and the RMS error of assembly power within 1.34% for all the cycles of all reactors.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.322-327
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• 2001

The internal leaks of RCS pressure boundary valves may cause thermal fatigue crack because of the TASCS in RCS branch line. After experienced unisolable piping failures in several PWR plants, many studies have peformed to understand these phenomena and various methods were applied to ensure the structural integrity of piping. In this paper, the cause of unisolable piping failures and the alternatives to prevent recurrence of failure were reviewed. Also, the severity of piping failure including susceptibility of valve leaks was evaluated for the Westinghouse 2-loop plant. The length of turbulent penetration on RHR inlet piping was measured and, thermal fluid analysis and fatigue analysis was performed for this piping. As a means of ensuring the structural integrity, temperature monitoring and specialized UT and other alternatives were compared for the further application.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2003.05a
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• pp.473-483
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• 2003

본 논문은 안전해석 등에 사용되는 RETRAN-3D 등 최적해석 코드를 기반으로 하면서도 복잡한 하드웨어 없이 간편한 GUI (Graphic User Interface)를 이용하여 광범위한 발전소 과도상태를 해석하기 위한 다양한 기능을 통해 시뮬레이션 조작을 쉽게 할 수 있는 웨스팅하우스형 950MW급 최적 원전운전분석기 (Nuclear Plant Analyzer)를 다루고자 한다. WH형 950MW 원전 운전최적분석기는 기존의 단순한 Point Kinetics 모델이 아닌 정교한 3D 실시간 노심모델과 RETRAN 코드를 기반으로 하는 실시간 NSSS 열수력 모델 (ARTS)이 통합된 모델을 갖추고 있으며, 해당형식발전소 (WH 3 Loop PWR Plant : 고리 3,4호기, 영광1,2호기 원전)의 여러 가지 과도사고를 실시간으로 정상, 비정상, 비상운전 등으로 모의할 수 있도록 개발되었다. 모의결과 주요 과도 상태의 결과가 해석한 결과와 잘 일치하였으며, 해당형식 발전소 과도 분석이나 규제요원 훈련에 이용될 계획이다.