• 한국재료학회:학술대회논문집
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• 한국재료학회 2002년도 춘계 학술발표강연 및 논문개요집
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• pp.166-166
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• 2002

• 한국재료학회:학술대회논문집
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• 한국재료학회 2001년도 춘계학술발표강연 및 논문개요집
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• pp.167-167
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• 2001

• 한국재료학회:학술대회논문집
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• 한국재료학회 2001년도 추계 학술발표강연 및 논문개요집
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• pp.194-194
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• 2001

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 춘계공동학술발표회요약집
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• pp.253-253
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• 2002

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 추계학술발표대회 요약집
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• pp.237.2-237.2
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• 2003

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.390-391
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• 2005

• 에너지공학
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• 제13권3호
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• pp.181-190
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• 2004

Computational models for analyzing the in-reactor behavior of metallic fuel pins under transient conditions in liquid-metal reactors are developed and implemented in the TRAMAC (TRAnsient thermo-Mechanical Analysis Code) for a metal fuel rod under transient operation conditions. Not only the basic models for a fuel rod performance but also some sub-models used for transient condition are installed in TRAMAC. Among the models, a fission gas release model, which takes the multi-bubble size distribution into account to characterize the lenticular bubble shape and the saturation condition on the grain boundary and the cladding deformation model have been developed based mainly on the existing models in the MAC-SIS code. Finally, cladding strains are calculated from the amount of thermal creep, irradiation creep, and irradiation swelling. The cladding strain model in TRAMAC predicts well the absolute magnitudes and gen-eral trends of their predictions compared with those of experimental data. TRAMAC results for the FH-1,2,6 pins are more conservative than experimental data and relatively reasonable than those of FPIN2 code. From the calculation results of TRAMAC, it is apparent that the code is capable of predicting fission gas release, and cladding deformation for LMR metal fuel finder transient operation conditions. The results show that in general, the predictions of TRAMAC agree well with the available irradiation data.

• Nuclear Engineering and Technology
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• 제43권3호
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• pp.301-308
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• 2011

The diametral creep of pressure tubes (PTs) in CANDU (CANada Deuterium Uranium) reactors is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of the heat transport system (HTS). PT diametral creep leads to diametral expansion, which affects the thermal hydraulic characteristics of the coolant channels and the critical heat flux (CHF). The CHF is a major parameter determining the critical channel power (CCP), which is used in the trip setpoint calculations of regional overpower protection (ROP) systems. Therefore, it is essential to predict PT diametral creep in CANDU reactors. PT diametral creep is caused mainly by fast neutron irradiation, temperature and applied stress. The objective of this study was to develop a bundle position-wise linear model (BPLM) to predict PT diametral creep employing previously measured PT diameters and HTS operating conditions. The linear model was optimized using a genetic algorithm and was devised based on a bundle position because it is expected that each bundle position in a PT channel has inherent characteristics. The proposed BPLM for predicting PT diametral creep was confirmed using the operating data of the Wolsung nuclear power plant in Korea. The linear model was able to predict PT diametral creep accurately.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2007년도 추계학술대회논문집
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• pp.281-286
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• 2007

Needs of fuel's performance evaluation for the dual-cooled fuel pellet (annular shape) necessitate the irradiation test in the test reactor. Irradiation test rig for the HARARO reactor, which is a special-purposed equipment used for material, irradiation and creep test, must satisfy the operational requirement on the hydraulic characteristics and structural integrity. In this paper, pressure drop and flow-induced vibration test for the newly developed non-instrumented test rig were carried out using FIVPET as a out-pile evaluation test. The test results show that the new test rig satisfy the HANARO operational requirement with sufficient margin. The spectral response characteristics of the flow-induced vibration of the test rid were also discussed.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2008년도 추계학술대회논문집
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• pp.226-229
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• 2008

To investigate an in-pile behavior of the newly developed DUO fuel pellet, the irradiation test will be carried out in the domestic test reactor. Irradiation test capsule for the HANARO reactor, which is a specially designed equipment used for material, irradiation and creep test, must satisfy the operational requirement on the hydraulic characteristics and structural integrity. In this study, a pressure drop, a flow-induced vibration and a short-term endurance test for the newly developed non-instrumented test capsule were carried out using FIVPET as a out-pile evaluation test. The test results show that the new test rig satisfy the HANARO operational requirement with sufficient margin.