• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2013년도 춘계학술대회 논문집
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• pp.1297-1298
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• 2013

• 한국재료학회지
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• 제4권4호
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• pp.445-452
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• 1994

가압 경수로 핵연료의 중성자 조사 조건에서 Zircaloy피복관의 3축방향으로의 변동거동은 집합도 계수에 따른 크립 이방성고 조사성장 이방성을 통하여 분석될 수 있다. 이러한 크립과 조사성장의 이방성이 Zircaloy피복관의 각 축방향 변형율에 미치는 영향을 평가할 수 있는 방법론이 제시되었다. 연소 후 측정된 KOFA Zircaloy-4피복관의 변형율과 핵연료 성능 분석 코드의예측치를 토대로 하여 각 축방향 변형율을 계산한 결과 KOFA Aircaloy-4 피복관의 원주방향 변형은 크립에 의해 주로 일어난 반면, 피복관의 길이방향 변형은 조사성장에 의하여 일어났으나 낮은 조사량에서는 크립의 영향도 상당히 큰것으로 나타났다.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.2041-2046
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• 2020

The present article investigates the influence of hydrogen concentration on the creep performance of cold-worked stress-relieved unirradiated Zircaloy-4 cladding tube using nanoindentation technique. The as-received Zircaloy-4 tube is hydrided to the concentrations of 600 ppm and 900 ppm using gaseous hydrogen charging method. Constant load indentation creep tests are performed for a dwell period of 600 s in the temperature range of 300℃-500 ℃ at 1000 μN, 2000 μN, and 3000 μN. The impact of hydrogen is evaluated in terms of steady state power law creep exponent and activation energy. The power law creep exponent decreases with increase in hydrogen concentration, however, it remains fairly constant with increase in temperature up to 500 ℃. Moreover, activation energy too decreases significantly with increase in hydrogen concentration. The mean stress exponent and activation energy are found to be 3.58 and 28.67 kJ/mol, respectively, for as-received sample.

• Nuclear Engineering and Technology
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• 제41권2호
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• pp.143-148
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• 2009

The increasing power and other demands on PWR fuel is leading to a demand for cladding that has low corrosion but that should also be robust with regard to mechanical behavior, impact of the irradiation environment and the coolant chemistry. The Optimized $ZIRLO^{TM}$ cladding is an evolutionary development of $ZIRLO^{TM}$ taking advantage of the long experience of the ZIRLO cladding but has significantly improved corrosion behavior. Recently, operation of Optimized ZIRLO to above 73 kWd/kgU has shown a reduction of the corrosion of almost 50%.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 추계학술발표회요약집
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• pp.256-256
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• 1998

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3723-3740
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• 2021

The harsh conditions in the reactor affect the thermal and mechanical performance of the fuel plate heavily. Some in-pile behaviors, like fission-induced swelling, can cause a large deformation of fuel plate at very high burnup, which may even disturb the flow of coolant. In this research, the emphasis is put on the thermal expansion, fission-induced swelling, interaction layer (IL) growth, creep of the fuel meat, and plasticity of the cladding for the U₃Si₂/Al dispersion fuel plate. A detailed model of the fuel meat swelling is developed. Taking these in-pile behaviors into consideration, the three-dimensional large deformation incremental constitutive relations and stress update algorithms have been developed to study its thermal-mechanical performance under normal conditions using Abaqus. Results have shown that IL can effectively decrease the thermal conductivity of fuel meat. The high Mises stress region mainly locates at the interface between fuel meat and cladding, especially around the side edge of the interface. With irradiation time increasing, the stress in the fuel plate gets larger resulting from the growth of fuel meat swelling but then decreases under the effect of creep deformation. For the cladding, plasticity deformation does not occur within the irradiation time.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.1031-1043
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• 2017

A structural model for stress distributions of coated Zircaloy subjected to realistic incore pressure difference, thermal expansion, irradiation-induced axial growth, and creep has been developed in this study. In normal operation, the structural integrity of coating layers is anticipated to be significantly challenged with increasing burnup. Strain mismatch between the zircaloy and the coated layer, due to their different irradiation-induced axial growth, and creep deformation are found to be the most dominant causes of stress. This study suggests that the compatibility of the high temperature irradiation-induced strains (axial growth and creep) between zircaloy and the coating layer and the capability to undergo plastic strain should be taken as key metrics, along with the traditional focus on chemical protectiveness.

• 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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.423-424
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• 2005

The performance of HANA claddings was evaluated in out-of-pile conditions. All the performance test results revealed that HANA claddings were superior to the reference claddings such as Zircaloy-4 and A-cladding. Corrosion resistance was improved by 60 to 70% compared to the commercial claddings. Creep, burst, tensile, LOCA, wear and microstructural properties were shown to be as good as the commercial claddings.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 추계학술발표대회 요약집
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• pp.361.2-361.2
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• 2003