• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.213-213
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• 2002

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.29-30
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• 2005

• Nuclear Engineering and Technology
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• 제38권6호
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• pp.535-550
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• 2006

The objectives of this paper are twofold to summarize the new findings and confirmed results from the Phebus FPT-1 experimental data and to report useful information to MELCOR users regarding the better use of MELCOR. For the core damage behavior, the early stage of a melt progression was predicted well; however, the late phase models, concerned with fuel dissolution, oxide cladding failure, fuel slumping, rubble debris heat up, effects of burn-up fuel, and so on, still showed limitations in MELCOR. For the fission product behavior, the comparison showed unexpected phenomena, various limitations, unresolved issues, and even absence of models. The issues summarized in this study have revealed the main areas where our endeavors need to be intensified in order to improve our understanding of severe accident phenomena. From the analysis of the Phebus FPT-1 test results, not only new core damage features, such as foaming or core expansion, but also possible new fission product release patterns due to effects from a high burn-up fuel have raised alternative challenging phenomena that should be solved in the next severe accident research phase.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.484-488
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• 1997

Fission product release (FPR) assessment for 100% reactor outlet header (ROH) break in CANDU reactor loaded with CANFLEX-NU fuel bundles has been performed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The fuel failure thresholds for the CANFLEX and standard bundle elements are very similar. All the sheaths at the corresponding fuel failure thresholds for the CANFLEX and standard bundles fail due to the significant cracks in the surface oxide, except those for the CANFLEX inner element at burnups of 220 to 240 MW.h/kg(U), which fail due to the excessive diametral strain. The fuel failure analysis predicts that the number of failed fuel elements for the CANFLEX bundle case is none, while that for the standard bundle case is 1827. The total (gap plus bound) I-131 releases for the CANFLEX and standard bundles are none and 5889 TBq, respectively The significant reduction of the number of failed fuel elements and FPR for the CABFKEX fuel bundle is attributed to the lower linear power of the CANFLEX fuel bundle compared with the standard fuel bundle.

• Nuclear Engineering and Technology
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• 제48권4호
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• pp.870-880
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• 2016

A steam generator tube rupture (SGTR) event with a stuck-open safety relief valve constitutes one of the most serious accident sequences in pressurized water reactors (PWRs) because it may create an open path for radioactive aerosol release into the environment. The release may be mitigated by the deposition of fission product particles on a steam generator's (SG's) dry tubes and structures or by scrubbing in the secondary coolant. However, the absence of empirical data, the complexity of the geometry, and the controlling processes have, until recently, made any quantification of retention difficult to justify. As a result, past risk assessment studies typically took little or no credit for aerosol retention in SGTR sequences. To provide these missing data, the Paul Scherrer Institute (PSI) initiated the Aerosol Trapping In Steam GeneraTor (ARTIST) Project, which aimed to thoroughly investigate various aspects of aerosol removal in the secondary side of a breached steam generator. Between 2003 and 2011, the PSI has led the ARTIST Project, which involved intense collaboration between nearly 20 international partners. This summary paper presents key findings of experimental and analytical work conducted at the PSI within the ARTIST program.

• Nuclear Engineering and Technology
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• 제31권1호
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• pp.88-103
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• 1999

The purposes of this study are to understand the severe accident phenomena, to establish the simulation method for the experimental test, and to assess the current models in MELCOR for future improvement. This paper presents the results of the PHEBUS FPT1 post test analysis using MELCOR computer code, version 1.8.4. The entire PHEBUS facility has been modeled; the core, the primary circuit including the steam generator, and the containment vessel. Both the thermal hydraulic and the fission product behavior have been investigated. The code simulation results of the thermal hydraulic behavior show good agreement with the experimental data, The fission product release and transport are calculated using the CORSOR models in MELCOR code and the results will be compared with the experiment when the experimental data are available.

• Nuclear Engineering and Technology
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• 제50권7호
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• pp.1148-1153
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• 2018

The Leaking Fuel Experiment test facility was designed to simulate the activity release from spent leaking fuel rods under steady state and transient conditions in the spent fuel pool. The experimental rig included an electrically heated fuel rod with different defects and a cooling system. The fission product transport was simulated by potassium-chloride. The conductivity changes of the water in the cooling system were measured to provide information about the amount of released solution. Defects of different sizes and positions were applied, together with a wide range of rod powers to simulate decay heat. The produced data can be used for predicting the activity release from leaking fuel under storage conditions and for the interpretation of fuel examination procedures.

• Nuclear Engineering and Technology
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• 제19권3호
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• pp.180-185
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• 1987

월성로형 핵연료봉의 정상상태 노내거동 분석용 전산코드인 KAFEPA를 개발하였다. KAFEPA 전산코드는 같은 목적하에서 AECL에 의해 개발되었던 ELESIM 전산코드에 대응하지만, KAPEPA 전산코드는 ELESIM에 비해 보다 이론적이고 정확하게 예측하는 계산모형들, 즉, 핵분열기체방출 모형, 노내고밀화모형 및 중성자속 감소 계산모형들을 내포하고 있다. KAFEPA 전산코드는 핵분열생성물 기체 방출에 대한 22개 노내 실험자료에 그 예측치를 비교함으로써 검증되었다. KAFEPA 전산코드에 의한 예측치는 상기 실험자료와 잘 일치하였다.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2702-2713
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• 2022

A steam generator tube rupture accompanying a core damage may cause the fission product to be released to environment bypassing the containment. In such an accident, the steam generator is the major path of the radioactive aerosol release. AEOLUS facility, the scaled-down model of Korean type steam generator, was built to examine the aerosol removal in the steam generator during the steam generator tube rupture accident. Integral and separate effect tests were performed with the facility for the dry and flooded conditions, and the decontamination factors were presented for different tube configurations and submergences. The dry test results were compared with the existing test results and with the analyses to investigate the aerosol retention physics by the tube bundle, with respect to the particle size and the bundle geometry. In the flooded tests, the effect of submergence were shown and the retention in the jet injection region were presented with respect to the Stokes number. The test results are planned to be used to constitute the aerosol retention model, specifically applicable for the analysis of the steam generator tube rupture accident in Korean nuclear power plants to evaluate realistic fission product behavior.

• Nuclear Engineering and Technology
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• 제16권3호
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• pp.169-179
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• 1984

FIPREL 전산코드를 사용하여 원자로 냉각수 내의 핵분열 생성물에 의한 방사능을 분석함으로써 PWR의 운전시에 발생하는 핵연료 피복관 파손을 평가할 수 있는 효과적인 절차를 모색하였다. 이 코드를 이용하여 핵연료의 농축도, 연소도, 가동온도 및 갭유출계수의 크기로 정량화되는 실제적 파손 크기등의 물리적 파라미터에 대해서 핵분열 생성물의 방사능이 나타내는 민감도에 대한 방대한 계산을 실시하였으며 그 결과는 PROFIP방법에 의한 것과 대체적으로 일치한다. 노출 우라늄이 존재하는 경우에는 옥소보다도 화학적으로 더 안정된 핵종간의 방사능비에 근거하여 반복계산을 실시함으로써 파손된 핵연료 봉에서 유출된 방사능만을 분리해 낸다. 개발된 전산코드로 파손 핵연료봉의 선형출력 밀도, 갯수, 실제적 파손 크기 및 노출우라늄의 질량등을 계산할 수 있다. 고리 1호기의 4주기에 걸친 운전 경험을 이 모텔에 의해 분석한 결과에 의하면 본 모델은 원자력발전소 정상운전시 핵연료봉의 상태를 감시·평가하는데 아주 적합한 것으로 판명되었다.