• Nuclear Engineering and Technology
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• v.37 no.4
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• pp.309-316
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• 2005

The contents of a new book currently in preparation are described. The dearth of books in the field of nuclear materials has left both students in nuclear materials classes and professionals in the same field without a resource for the broad fundamentals of this important sub-discipline of nuclear engineering. The new book is devoted entirely to materials problems in the core of light-water reactors, from the pressure vessel into the fuel. Key topics deal with the $UO_2$ fuel, Zircaloy cladding, stainless steel, and of course, water. The restriction to LWR materials does not mean a short monograph; the enormous quantity of experimental and theoretical work over the past 50 years on these materials presents a challenge of culling the most important features and explaining them in the simplest quantitative fashion. Moreover, LWRs will probably be the sole instrument of the return of nuclear energy in electric power production for the next decade or so. By that time, a new book will be needed.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.139-146
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• 1997

Zircaloy-4의 수증기 산화와 이에 따른 수소침투의 압력에 대한 영향을 평가하기 위해 pre-transition과 post-transition의 영역에서 1~103 기압의 압력 범위에서 실험을 수행하였다. 그리고 시편의 edge부분에서의 산화율 및 수소침투 가속화 현상을 알아보기 위해 시편의 edge 분율에 따른 산화율 및 수소침투량 실험을 압력영향과 함께 수행하였다. 또 steam corrosion과 waterside corrosion의 비교를 위해 산화율에 따른 수소침투를 평가하였다. 잠정적인 결과로서 pre-transition 영역, 즉, 37$0^{\circ}C$, 72시간에서 103기압에서의 산화가 1 기압에서의 산화보다 약 50% 증가된 값을 가졌고,post-transition 영역, 즉, $700^{\circ}C$, 210분에서는 최고 150%의 산화 가속화를 관찰할 수 있었으며 수소 침투량 역시 산화가 가속화된 만큼 증가하였다. 그리고 압력이 증가함에 따라 산화율이 점진적으로 증가함을 pre-transition영역과 post-transition영역에서 관찰할 수 있었다. 시편의 edge 분율에 따른 산화율의 변화에 대해서는 37$0^{\circ}C$, 72시간의 경우 산화량이 적어 별다른 영향을 관찰할 수 없었으나, $700^{\circ}C$, 210분에서는 시편의 표면적에 대한 edge의 비율이 증가할수록 산화율이 증가하고 있음을 볼 수 있다. 하지만 기존의 논문들에서 주장하고 있는 뚜렷한 edge의 영향을 관측하기는 어려웠다.

• Nuclear Engineering and Technology
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• v.53 no.7
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• pp.2295-2303
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• 2021

To investigate wall-thinning impact on axial load resistance of Zircaloy-4 cladding rods after a LOCA transient, axial tensile samples have been machined on as-received tubes with reduced thicknesses between 370 and 580 ㎛. After high temperature oxidation under steam at 1200 ℃ with measured ECR ranging from 10 to 18% and water quenching, machined samples were axially loaded until fracture. These tests were modeled using a fracture mechanics approach developed in a previous study. Fracture stresses are rather well predicted. However, the slightly lower fracture stress observed for wall-thinned samples is not anticipated by this modeling approach. The results from this study confirm that characterizing the axial load resistance using semi-integral tests including the creep and burst phases was the best option to obtain accurate axial strengths describing accurately the influence of wall-thinning at burst region.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1250-1258
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• 2021

In this paper, the mechanical performance of cracked surface-coated Zircaloy cladding, which has different coating materials, coating thicknesses and initial crack lengths, has been investigated. By analyzing the stress field near the crack tip, the safety zone range of initial crack length has been decided. In order to determine whether the crack can propagate along the radial (r) or axial (z) directions, the energy release rate has been calculated. By comparing the energy release rate with fracture toughness of materials, we can divide the initial crack lengths into three zones: safety zone, discussion zone and danger zone. The results show that Cr is suitable coating material for the cladding with a thin coating while Fe-Cr-Al have a better fracture mechanical performance in the cladding with thick coating. The Si-coated and SiC-coated claddings are suitable for reactors with low power fuel elements. Conclusions in this paper can provide reference and guidance for the cladding design of nuclear fuel elements.

• Proceedings of the Korean Institute of Surface Engineering Conference
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• 2006.04a
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• pp.93-94
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• 2006

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.230-230
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• 1998

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.246.1-246
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.241-241
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.242-242
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.364.2-364.2
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• 2003