• 에너지공학
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• 제12권3호
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• pp.177-183
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• 2003

본 논문은 방사능조사의 조건하에서의 설계조건을 만족해야 하는 fuel rod의 설계응력값을 검증하는데 그 목적이 있다. 이 경우에, 안전성에 있어서 열에 의한 영향이 가장 주된 고려조건이 된다. 그러나 이러한 열영향이 고려된 해석은 구조물의 안전성해석에서 비교적 간단히 해결되는 문제가 아니다. 여기서는 이론적 해석을 통한 접근방식으로 보수적인 관점에서 fuel rod의 설계에 적용되는 설계 응력값을 검증하고자 하였다. 추후에 시도하는 fuel rod의 설계에 있어서 본 해석방법을 이용하면 안전설계의 검증을 위한 이론적 접근방법의 하나로 이용할 수 있을 것으로 사료된다.

• 한국소음진동공학회논문집
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• 제20권3호
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• pp.215-222
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• 2010

Long and slender body, like a fuel rod, oscillating in axial flow can be unstabilized even by the small cross flow which can be activated by the flow mixer or turbulent generator. It is important to include these effects of flow disturbance in dynamic stability analysis of nuclear fuel rod. This work shows how eigen frequency of a multi-span fuel rod can be changed by the swirl flow, which is discretely generated by a flow mixer. By solving a state-space form of the eigenvalue equation for a multi-span fuel rod system, the critical velocity at which a fuel rod becomes unstable was calculated. Based on the simulation results, we evaluated how stability of a multi-spanned nuclear fuel rod with mixing vanes can be affected by the coolant flow in an operating reactor core.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2008년도 춘계학술대회논문집
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• pp.641-646
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• 2008

Long and slender body with or without flexible supports under severe operating condition can be unstabilized even by the small cross flow. Turbulent flow mixer, which actually increases thermal-hydraulic performance of the nuclear fuel by boosting turbulence, disturbs the flow field around the fuel rod and affects dynamic behavior of the nuclear fuel rods. Few studies on this problem can be found in the literature because these effects depend on the specific natures of the support and the design of the system. This work shows how the dynamics of a multi-span fuel rod can be affected by the turbulent flow, which is discretely activated by a flow mixer. By solving a state-space form of the eigenvalue equation for a multi-span fuel rod system, the critical velocity at which a fuel rod becomes unstable was established. Based on the simulation results, we evaluated how stability of a multi-spanned nuclear fuel rod with mixing vanes can be affected by the coolant flow in an operating reactor core.

• Nuclear Engineering and Technology
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• 제33권1호
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• pp.73-82
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• 2001

Pressurized water reactor(PWR) fuel rods. which are continuously supported by a spring system called a spacer grid(SG), are exposed to reactor coolant at a flow velocity of up to 6-8 m/s. It is known that the vibration of 3 fuel rod is generated by the coolant flow, a so-called flow-induced-vibration(FIV), and the relative motion induced by the FIV between the fuel rod and the SG can wear away the surface of the fuel rod, which occasionally leads to its fretting failure. It is, therefore, important to understand the vibration characteristics of the fuel rod and reflect that in its design. In this paper, vibration analyses of the fuel rod with two different SGs were performed using both analytical and experimental methods. Updating of the finite element(FE) model using the measured data was performed in order to enhance confidence in the FE model of fuel rods supported by an SG. It was found that the modal parameters are very sensitive to the spring constant of the SG.

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.1079-1092
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• 2020

A parametric study of several parameters relevant to design safety on the spent nuclear fuel (SNF) rod response under a drop accident is presented. In the view of the complexity of interactions between the independent safety-related parameters, a factorial design of experiment is employed as an efficient method to investigate the main effects and the interactions between them. A detailed single full-length fuel rod is used with consideration of post-irradiated fuel conditions under horizontal and vertical free-drops onto an unyielding surface using finite-element analysis. Critical drop heights and critical g-loads that yield the threshold plastic strain in the cladding are numerically estimated to evaluate the fuel rod structural resistance to impact load. The combinatory effects of four uncertain parameters (pellet-cladding interfacial bonding, material properties, spacer grid stiffness, rod internal pressure) and the interactions between them on the fuel rod response are investigated. The principal finding of this research showed that the effects of above-mentioned parameters on the load-carrying capacity of fuel rod are significantly different. This study could help to prioritize the importance of data in managing and studying the structural integrity of the SNF.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2002년도 춘계학술대회 논문집
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• pp.684-687
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• 2002

The authors have developed two devices which cuts the spend fuel rod to an optimal size and extracts fuel pellet from the pieces of cut fuel rods. These devices are so important to reduce radioactive wastes that some advanced countries developed their own methods and devices. The authors have benchmarked from these methods and devices. For spent fuel rod cutting, the tube cutting method has been chosen. some mechanical properties of the fuel tube and pellet has been carefully considered for an optimal cutting size. For fuel pellet extraction, a mechanically extracting method has been adopted. The existing chemical method have turned out to be inappropriate because it produced large amount of radioactive wastes, in spite of its high fuel recovery characteristics. The developed method has an advantage that it can be applied to other fuel rods that have different shapes and sizes. The two devices are set up and operated in the hot cell where people can not go in, so that the devices have been designed to be controlled remotely and modulated for easy maintenance. And the performance of the devices has been tested by using simulated fuel rod. From the experimental results, the devices are supposed to be useful for reducing radioactive wastes.

• Nuclear Engineering and Technology
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• 제30권3호
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• pp.273-286
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• 1998

As there has arisen a concern that failure of the high burnup fuel under the reactivity-insertion accident(RIA) may occur at the energy lower than the expected, fuel behavior under the rod ejection accident in a typical Westinghouse-designed 950 MWe PWR was analyzed by using the three dimensional nodal transient neutronics code, PANBOX2 and the transient fuel rod performance analysis code, FRAP-T6. Fuel failure criteria versus the burnup was conservatively derived taking into account available test data and the possible fuel failure mechanisms. The high burnup and longer cycle length fuel loading scheme of a peak rod turnup of 68 MWD/kgU was selected for the analysis. Except three dimensional core neutronics calculation, the analysis used the same core conditions and assumptions as the conventional zero dimensional analysis. Results of three dimensional analysis showed that the peak fuel enthalpy during the rod ejection accident is less than one third of that calculated by the conventional zero dimensional analysis methodology and the fraction of fuel failure in the core is less than 4 %. Therefore, it can be said that the current design limit of less than 10 percent fuel failure and maintaining the core coolable geometry would be adequately satisfied under the rod ejection accident, even though the conservative fuel failure criteria derived from the test data are applied.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 추계학술대회논문집B
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• pp.3-8
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• 2001

For the successful design of nuclear reactor, it is very important to investigate thermal-hydraulic characteristics of fuel rod bundle. Fluid flow and heat transfer in the non-circular cross-section of nuclear fuel rod bundle are different from those found in common circular tube. And complex three dimensional flow including secondary and vortex flow, is formed around the bundles. The purpose of this research is to examine how geometries and flow conditions affect heat transfer in fuel rod bundle. Design data for nuclear fuel rod bundle and structure are surveyed, and $3{\times}3$ sub-channel model is adopted in this study. Computational results are compared with the heat transfer data measured by naphthalene sublimation method, and numerical analysis and evaluation are performed at various design conditions and flow conditions.

• Journal of Radiation Protection and Research
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• 제34권3호
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• pp.144-150
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• 2009

This study examines how much the radiation dose rate around it varies if a crack occurs on the spent nuclear fuel rod. The spent nuclear fuel rod to be examined is that of Kori unit 3&4. The source terms are evaluated using the ORIGEN-ARP that is part of the version 5.1 of the SCALE package. The radiation dose rate is assessed using the TORT. To check if the structure of a fuel rod is appropriately modeled in the TORT calculation, the calculation results by the TORT are compared with those by the ANISN for the same case. From the code simulation, it is known that if a crack occurs on the spent nuclear fuel rod, the neutron dose rate varies depending on what material is the crack filled with, but the gamma dose rate varies irrespective of type of the material that the crack is filled with.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2003년도 춘계학술대회논문집
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• pp.683-688
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• 2003

Vibration characteristics of a fuel rod to be used in KALIMER(Korean Advanced LIquid MEtal Reactor) have been estimated through 3-dimensional finite element analysis and verified by experiment. The fundamental natural frequencies are found to be 6㎐ in air and 2.5㎐ in water. respectively. It has been found that in-water natural frequencies of the fuel rod are lower than in-air ones due to the added mass effect of the fluid filled inside the outer cylinder and they further decreases as the gap between the fuel rod and the outer cylinder increases, namely the added mass effect increases as the gap increases(maximum 54%). It has been also shown that the mass of the wire wrap axially coiled around the fuel rod do not affect the natural frequencies.