• 방사성폐기물학회지
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• 제19권2호
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• pp.215-223
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• 2021

The domestic CANDU nuclear power plants have been operated for a long time and various unforeseen spent fuel defects have been discovered. As the spent fuel defects are important factors in the safety of the nuclear power plant, a study on the analysis of the spent fuel defects to prevent their recurrence is necessary. However, in cases where the fuel rods inside the fuel assembly are defected, it is difficult to dismantle the fuel assembly owing to their welded structure and the facility conditions of the plant. Therefore, it is impossible to analyze the spent fuel defect because it is difficult to visually check the shape of the fuel defect. To resolve these problems, an analysis technology that can predict the number of defected fuel rods and defect size was developed. In this study, we developed a methodology for investigating the root cause of spent fuel defects using a database of the earlier fuel defects in the plants. It is anticipated that in the future this analysis technology will be applied when spent fuel defects occur.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1333-1344
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• 2019

An integrated approach of model simplification for high burnup spent nuclear fuel is proposed based on material calibration using optimization. The spent fuel rods are simplified into a beam with a homogenous isotropic material. The proposed approach of model simplification is applied to fuel rods with two kinds of interfacial configurations between the fuel pellets and cladding. The differences among the generated models and the effects of interfacial bonding efficiency are discussed. The strategy of model simplification adopted in this work is to force the simplified beam model of spent fuel rods to possess the same compliance and failure characteristics under critical loads as those that result in the failure of detailed fuel rod models. It is envisioned that the simplified model would enable the assessment of fuel rod failure through an assembly-level analysis, without resorting to a refined model for an individual fuel rod. The effective material properties of the simplified beam model were successfully identified using the integrated optimization process. The feasibility of using the developed simplified beam models in dynamic impact simulations for a horizontal drop condition is examined, and discussions are provided.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2005년도 창립60주년 기념 추계 학술대회
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• pp.219.2-219.2
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• 2005

• 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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• 제54권11호
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• pp.4030-4048
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• 2022

In this study, road transportation tests were conducted with surrogate fuel assemblies under normal conditions of transport to evaluate the vibration and shock load characteristics of spent nuclear fuel (SNF). The overall test data analysis was conducted based on the measured acceleration and strain data obtained from the speed bump, lane-change, deceleration, obstacle avoidance, and circular tests. Furthermore, representative shock response spectrums and power spectral densities of each test mode were acquired. Amplification or attenuation characteristics were investigated according to the load transfer path. The load attenuated significantly as it transferred from the trailer to the cask. By contrast, the load amplified as it transferred from the cask to the surrogate SNF assembly. The fuel loading location on the cask disk assembly did not exhibit a significant influence on the strain measured from the fuel rods. The principal strain was in the vertical direction, and relatively large strain values were obtained in spans with large spacing between spacer grids. The influence of the lateral location of fuel rods was also investigated. The fuel rods located at the side exhibited relatively large strain values than those located at the center. Based on the strain data obtained from the test results, a hypothetical road transportation scenario was established. A fatigue evaluation of the SNF rod was performed based on this scenario. The evaluation results indicate that no fatigue damage occurred on the fuel rods.

• 방사성폐기물학회지
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• 제14권2호
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• pp.149-156
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• 2016

저장조에 위치한 사용후핵연료는 가혹한 원자로 조건에 의해 구조적 건전성이 와해되므로 외력에 취약하다. 따라서 운반 및 취급 중 사고 상황이 고려되어야 한다. 극단적인 경우, 핵연료 취급 중 사고로 인해 핵연료 저장조에서 핵연료집합체 낙하가 발생할 수 있다. 이러한 사고 상황 하에서 연료봉 파손 등을 평가하기 위해서 수조에 충돌할 때 발생하는 충돌력을 분석할 필요가 있다. 이는 핵연료가 수조 바닥에 충돌할 때의 속도를 입력으로 하여 평가될 수 있다. 연료봉이 핵연료 중량 및 부피의 대부분을 차지하고 있으므로, 연료봉 다발은 수중 항력을 예측하는데 중요한 역할을 한다고 볼 수 있다. 본 연구에서는 $3{\times}3$ 의 짧은 연료봉 다발을 모델로 사용하여 수중에서 낙하할 때 받는 수력을 계산하였고, 이를 전산모사와의 비교를 통하여 검증하였다. 본 방법론을 사용후핵연료 건전성 평가에 적용할 수 있을 것으로 기대된다.

• 비파괴검사학회지
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• 제22권1호
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• pp.9-13
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• 2002

수중에서 사용후 핵연료 제원측정 시험의 효율성을 높이고 측정오차를 줄이기 위하여 수중 영상측정방법을 개발하였다. 이 시스템의 모의 핵연료봉 직경 및 길이 측정치는 실제값 기준으로 할 때, 각각 $-0.24{\pm}0.03mm,\;0.34{\pm}0.06mm$이고 측정 최대오차는 각각 -0.3mm 및 0.4mm이내였다. 실제 사용후핵연료에 대한 수중 제원측정결과 고리원자력 2호기에서 2주기 동안 연소한 핵연료 집합체 J44의 핵연료봉 직경은 설계치 기준으로 할 때 핵연료봉 상 하단부 직경은 2.0%, 중앙부의 직경은 3.0% 정도 감소하였으나 핵연료봉의 길이는 0.4% 정도 신장하였다. 고리원자력 1호기에서 3주기 동안 연소한 핵연료 집합체 F02의 핵연료봉의 직경 및 길이는 핵연료 집합체 J44의 결과와 비슷한 경향을 나타내었다.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 2001년도 ICCAS
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• pp.123.3-123
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• 2001

To recycle the uranium resources in the spent nuclear fuels, all the fuel rods are extracted from the spent fuel assemblies. The remaining components of the spent fuel assembly after extracting all the rods, so called a NFBC(Non-Fuel Bearing Components), should be compacted to minimize the waste volume. To this present, KAERI (Korea Atomic Research Institute) has developed he NFBC compactor by introducing a new concept of cutting and compaction, In this paper, to achieve he maximum compaction ration of the NFBC volume while reducing compactor size, an fuzzy controller, which determines the reference force of the compactor, is proposed with using he fuzzy-inference.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.11-18
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• 2022

In this research, curium oxide (CmO₂) is studied as fuel for VVER-1200 reactor to get an attention to its energy value and possibilities. For this purpose, CmO₂ is used in fuel rods or integrated burnable absorber (IBA) rods with and without UO₂ and then compared with the conventional fuel assembly of VVER-1200 reactor. It is burned to 60 GWd/t by using SRAC-2006 code and JENDL-4.0 data library. From these studies, it is found that CmO₂ is competent like UO₂ as a fuel due to higher fission cross-section of ²⁴³Cm and ²⁴⁵Cm isotopes and neutron capture cross-section of ²⁴⁴Cm and ²⁴⁶Cm isotopes. As a result, when some or all of the UO₂ of fuel rods or IBA rods are replaced by CmO₂, we get a similar k-inf like the reference even with lower enrichment UO₂ fuels. These studies show that the use of CmO₂ as IBA rods is more effective than the fuel rods considering the initially loaded amount, power peaking factor (PPF), fuel temperature and void coefficient, and the quality of spent fuel. From a detailed study, 3% CmO₂ with inert material ZrO₂ in IBA rods are recommended for the VVER-1200 reactor assembly from the once through concept.

• 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.