• 한국콘텐츠학회논문지
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• 제13권10호
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• pp.46-53
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• 2013

원전 핵 연료봉에 대한 검사는 고준위 방사능의 위험으로부터 원격에서 이루어져야 하며, 또한 정확하게 이루어져야 한다. 현재 원전 핵 연료봉에 대한 검사는 방사능의 위험으로 인해 카메라로 핵 연료봉을 촬영하여 녹화된 동영상을 원격에서 보고 판단하는 형태로 운영되고 있다. 본 연구는 원자력 발전소에서 사용되는 핵 연료봉에 대한 상태 검사를 영상처리를 통하여 구현한 것이다. 핵 연료봉은 여러 개를 다발로 묶어 사용하도록 되어있는데, 이를 검사하기위한 영상처리 기법으로는 에지 검출 기법이 유용하다. 핵 연료봉 에지 검출을 위해 DoG 기법에 임계값 처리를 추가하는 방법을 제안한다. 이것은 DoG를 최적화한 새로운 기법으로, DoG와 임계값을 듀얼로 처리하는 방법이다. 이 방법으로 핵 연료봉 에지를 검출한 후에, 핵 연료봉 검사 알고리즘을 수행하여 핵 연료봉 이상 유무를 판단한다. 이러한 방법을 통해 미니어처로 제작된 핵 연료봉 검사시스템에 실험한 결과 우수한 특성을 확인하였다. 본 연구를 통해 원전 핵 연료봉 상태 검사를 보다 쉽고 안전하게 시행할 수 있을 것으로 생각된다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2019년도 춘계학술논문요약집
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• pp.94-95
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• 2019

• Nuclear Engineering and Technology
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• 제52권6호
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• pp.1099-1109
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• 2020

The performance of gadolinium burnable absorber (GdBA) for reactivity control in UO₂ and (U, Th)O₂ fuels and its impact on spent fuel characteristics was performed. Five fuel assemblies: one without GdBA fuel rod and four each containing 16, 24, 34 and 44 GdBA fuel rods in both fuels were investigated. Reactivity swing in all the FAs with GdBA rods in UO₂ fuel was higher than their counterparts with similar GdBA fuel rods in (U, Th)O₂ fuel. The excess reactivity in all FAs with (U, Th)O₂ fuel was higher than UO₂ fuel. At the end of single discharge burn-up (~ 49.64 GWd/tHM), the excess reactivity of (U, Th) O₂ fuel remained positive (16,000 pcm) while UO₂ fuel shows a negative value (-6,000 pcm), which suggest a longer discharge burn-up in (U, Th)O₂ fuel. The concentration of plutonium isotopes and minor actinides were significantly higher in UO₂ fuel than in (U, Th)O₂ fuel except for ²³⁶Np. However, the concentration of non-actinides (gadolinium and iodine isotopes) except for ¹³⁵Xe were respectively smaller in (U, Th)O₂ fuel than in UO₂ fuel but may be two times higher in (U, Th)O₂ fuel due to its potential longer discharge burn-up.

• Nuclear Engineering and Technology
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• 제39권1호
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• pp.85-94
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• 2007

The Korean nuclear fuel cycle was modeled by the dynamic analysis method, which was applied to the once-through and alternative fuel cycles. First, the once-through fuel cycle was analyzed based on the Korean nuclear power plant construction plan up to 2015 and a postulated nuclear demand growth rate of zero after 2015. Second, alternative fuel cycles including the direct use of spent pressurized water reactor fuel in Canada deuterium uranium reactors (DUPIC), a sodium-cooled fast reactor and an accelerator driven system were assessed and the results were compared with those of the once-through fuel cycle. The once-through fuel cycle calculation showed that the nuclear power demand would be 25 GWe and the amount of the spent fuel will be ${\sim}65000$ tons by 2100. The alternative fuel cycle analyses showed that the spent fuel inventory could be reduced by more than 30% and 90% through the DUPIC and fast reactor fuel cycles, respectively, when compared with the once-through fuel cycle. The results of this study indicate that both spent fuel and uranium resources can be effectively managed if alternative reactor systems are timely implemented along with the existing reactors.

• 방사성폐기물학회지
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• 제8권4호
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• pp.269-277
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• 2010

원자력의 중흥기를 맞이하여 민감한 핵연료주기 기술의 무분별한 확장을 억제하고자 다양한 핵연료주기 다자 방안이 제시되고 있다. 현재 원자력 공급국 위주의 핵연료주기 다자화가 추진되고 있는 실정에서 후행 핵연료주기 기술의 다자화 추진 추이를 파악하고자 사용후핵연료 공동 관리 시설에 대한 분석 결과를 검토하였다. 또한 후행 핵연료주기 연구개발 시설의 다자화를 제안하고 기대효과와 문제점을 검토한 후 이를 실현하기 위한 추진방안을 도출하였다.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.2195-2205
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• 2023

One of the salient nuclear fuel performance parameters for new fuel types under development is changes in fuel thickness. To test the new commercially fabricated U-10Mo monolithic plate-type fuel, an irradiation experiment was designed that consisted of multiple mini-plate capsules distributed within the Advanced Test Reactor (ATR) core, the mini-plate 1 (MP-1) experiment. Each capsule contains eight mini-plates that were either fueled or "dummy" plates. Fuel thickness changes within a fuel assembly can be characterized by measuring the gaps between the plates ultrasonically. The channel gap probe (CGP) system is designed to measure the gaps between the plates and will provide information that supports qualification of U-10Mo monolithic fuel. This study will discuss the design and the results from the use of a custom-designed CGP system for characterizing the gaps between mini-plates within the MP-1 capsules. To ensure accurate and repeatable data, acceptance and calibration procedures have been developed. Unfortunately, there is no "gold" standard measurement to compare to CGP measurements. An effort was made to use plate thickness obtained from post-irradiation measurements to derive channel gap estimates for comparison with the CGP characterization.

• Nuclear Engineering and Technology
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• 제53권6호
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• pp.1869-1877
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• 2021

This paper presents the neutron dose rate analysis of the new CONSTOR® RBMK-1500/M2 storage cask intended for the spent nuclear fuel storage at Ignalina Nuclear Power Plant in Lithuania. These casks are designed to be stored in a new "closed" type interim storage facility, with the capacity to store up to 202 CONSTOR® RBMK-1500/M2 casks. In 2016 y, the "hot trials" of this new facility were conducted and 10 CONSTOR® RBMK-1500/M2 casks loaded with the spent nuclear fuel were transported to the dedicated storage places in this facility. During "hot trials", the dose rate measurements of the CONSTOR® RBMK-1500/M2 casks were performed as the dose rate is one of the critical parameter to control and it must be below design (and safety) criteria. Therefore, having the actual data of the spent nuclear fuel characteristics, the neutron dose rate modeling of the CONSTOR® RBMK-1500/M2 cask loaded with this particular fuel was also performed. Neutron dose rate modeling was performed using MCNP 5 computer code with very detailed geometrical representation of the cask and the fuel. The obtained modeling results were compared with the measurement results and it was revealed, that modeling results are generally in good agreement with the measurements.

• Nuclear Engineering and Technology
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• 제43권4호
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• pp.329-334
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• 2011

Two conceptual flowsheets were developed for recycling used nuclear fuel. One flowsheet was developed for recycling used oxide nuclear fuel from light water reactors while the other was developed for recycling used metal fuel from fast spectrum reactors. Both flowsheets were developed from a set of design principles including efficient actinide recovery, nonproliferation, waste minimization and commercial viability. Process chemistry is discussed for each unit operation in the flowsheet.

• Nuclear Engineering and Technology
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• 제41권6호
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• pp.821-830
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• 2009

In this paper, modal testing and finite element modeling results to identify the modal parameters of a nuclear fuel rod as well as its cladding tube are discussed. A vertically standing full-size cladding tube and a fuel rod with lead pellets were used in the modal testing. As excessive flow-induced vibration causes a failure in fuel rods, such as fretting wear, the vibration level of fuel rods should be low enough to prevent failure of these components. Because vibration amplitude can be estimated based on the modal parameters, the dynamic characteristics must be determined during the design process. Therefore, finite element models are developed based on the test results. The effect of a lumped mass attached to a cladding tube model was identified during the finite element model optimization process. Unlike a cladding tube model, the density of a fuel rod with pellets cannot be determined in a straightforward manner because pellets do not move in the same phase with the cladding tube motion. The density of a fuel rod with lead pellets was determined by comparing natural frequency ratio between the cladding tube and the rod. Thus, an improved fuel rod finite element model was developed based on the updated cladding tube model and an estimated fuel rod density considering the lead pellets. It is shown that the entire pellet mass does not contribute to the fuel rod dynamics; rather, they are only partially responsible for the fuel rod dynamic behavior.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1977년도 제9회 학술발표회 초록
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• pp.112-112
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• 1977