• 한국압력기기공학회 논문집
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• 제13권2호
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• pp.75-83
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• 2017

In severe accident conditions of light water reactors, the loss of coolant may cause problems in integrity of zirconium fuel cladding. Under the condition of the loss of coolant, the zirconium fuel cladding can be exposed to high temperature steam and reacted with them by producing of hydrogen, which is caused by the failure in oxidation resistance of zirconium cladding materials during the loss of coolant accident scenarios. In order to avoid these problems, we develop a multi-metallic layered composite (MMLC) fuel cladding which compromises between the neutronic advantages of zirconium-based alloys and the accident-tolerance of non-zirconium-based metallic materials. Cold pilgering process is a common tube manufacturing process, which is complex material forming operation in highly non-steady state, where the materials undergo a long series of deformation resulting in both diameter and thickness reduction. During the cold pilgering process, MMLC claddings need to reduce the outside diameter and wall thickness. However, multi-layers of the tube are expected to occur different deformation processes because each layer has different mechanical properties. To improve the utilization of the pilgering process, 3-dimensional computational analyses have been made using a finite element modeling technique. We also analyze the dimensional change, strain and stress distribution at MMLC tube by considering the behavior of rolls such as stroke rate and feed rate.

• 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.

• 에너지공학
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• 제21권3호
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• pp.254-264
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• 2012

원자력발전소 중대사고시 노심용융물의 Rayleigh-Benard 자연대류 문제에 대한 예비실험으로 두 평판 사이의 거리, 측면벽의 유무 및 평판의 기하구조가 열전달에 미치는 영향에 대해 실험적 연구를 수행하였다. 열전달 실험을 대신하여 상사성의 원리를 이용한 황산-황산구리 수용액의 전기도금계를 물질전달계로 채택하였다. 실험은 $Ra_s$는 $1.06{\times}10^7{\sim}2.91{\times}10^{10}$의 범위에서 실험적 조건을 변화시켜가며 열전달을 측정하였다. 실험결과 단일 수평평판에서 측정한 열전달은 McAdams의 수평평판 자연대류 열전달 상관식과 일치하였고 두 평판에서 측정한 열전달은 Dropkin과 Somerscales, Globe와 Dropkin의 Rayleigh-Benard 자연대류 열전달 상관식과 매우 유사한 경향을 보였다. 두 평판 사이의 거리가 작을 경우 열전달이 높다가 거리가 증가하면 단일 수평평판에서의 자연대류 열전달과 같아졌다. 평판에 설치된 휜(Fin)은 열전달을 향상시켰다. 모든 경우에서 측면벽이 없는 경우의 열전달이 측면 벽이 있는 경우보다 항상 높았다.

• 한국구조물진단유지관리공학회 논문집
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• 제5권2호
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• pp.147-154
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• 2001

The prestressed concrete containment is one of the most important structures in nuclear power plants, which is required to prevent release of radioactive or hazardous effluents to the environment even in the case of a severe accident. Numerical analyses are carried out by using the ABAQUS finite element program to assess the ultimate pressure capacity of the Y prestressed concrete containment with light water reactor at design criteria condition and aging condition considering varied properties of time-dependant materials respectively. From the results, it is verified that the structural capacity of the Y prestressed concrete containment building under the present, aging condition is still robust. In addition, the parameter studies for the reduction of the ultimate pressure capacity of containment building according to the degradation levels of the main structural materials are carried out. The results show that when the degradations of each materials are considered as individual and combined forms, the influence is large in the order of tendon, rebar and concrete degradation, and tendon-rebar, tendon-concrete and rebar-concrete degradation respectively.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.218-227
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• 2016

Background: Steam explosions may occur in nuclear power plants by molten fuel-coolant interactions when the external reactor vessel cooling strategy fails. Since this phenomenon can threaten structural barriers as well as major components, extensive integrity assessment research is necessary to ensure their safety. Method: In this study, the influence of yield criteria was investigated to predict the failure of a reactor cavity under a typical postulated condition through detailed parametric finite element analyses. Further analyses using a geometrically simplified equivalent model with homogeneous concrete properties were also performed to examine its effectiveness as an alternative to the detailed reinforcement concrete model. Results: By comparing finite element analysis results such as cracking, crushing, stresses, and displacements, the Willam-Warnke model was derived for practical use, and failure criteria applicable to the reactor cavity under the severe accident condition were discussed. Conclusion: It was proved that the reactor cavity sustained its intended function as a barrier to avoid release of radioactive materials, irrespective of the different yield criteria that were adopted. In addition, from a conservative viewpoint, it seems possible to employ the simplified equivalent model to determine the damage extent and weakest points during the preliminary evaluation stage.

• Journal of Mechanical Science and Technology
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• 제18권6호
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• pp.990-1001
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• 2004

The retention of a molten pool vessel cooled by internal vessel reflooding and/or external vessel reactor cavity flooding has been considered as one of severe accident management strategies. The present numerical study investigates the effect of both internal and external vessel mixed cooling on an internally heated molten pool. The molten pool is confined in a hemispherical vessel with reference to the thermal behavior of the vessel wall. In this study, our numerical model used a scaled-down reactor vessel of a KSNP (Korea Standard Nuclear Power) reactor design of 1000 MWe (a Pressurized Water Reactor with a large and dry containment). Well-known temperature-dependent boiling heat transfer curves are applied to the internal and external vessel cooling boundaries. Radiative heat transfer has been considered in the case of dry internal vessel boundary condition. Computational results show that the external cooling vessel boundary conditions have better effectiveness than internal vessel cooling in the retention of the melt pool vessel failure.

• Structural Engineering and Mechanics
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• 제50권5호
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• pp.617-633
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• 2014

Because the elevated temperature degrades the mechanical properties of materials used in containments, the global behavior of containments subjected to the internal pressure under high temperature is remarkably different from that subjected to the internal pressure only. This paper concentrates on the nonlinear finite element analyses of the nuclear power plant containment structures, and the importance for the consideration of the elevated temperature effect has been emphasized because severe accident usually accompanies internal high pressure together with a high temperature increase. In addition to the consideration of nonlinear effects in the containment structure such as the tension stiffening and bond-slip effects, the change in material properties under elevated temperature is also taken into account. This paper, accordingly, focuses on the three-dimensional nonlinear analyses with thermal effects. Upon the comparison of experiment data with numerical results for the SNL 1/4 PCCV tested by internal pressure only, three-dimensional analyses for the same structure have been performed by considering internal pressure and temperature loadings designed for two kinds of severe accidents of Saturated Station Condition (SSC) and Station Black-out Scenario (SBO). Through the difference in the structural behavior of containment structures according to the addition of temperature loading, the importance of elevated temperature effect on the ultimate resisting capacity of PCCV has been emphasized.

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.925-936
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• 2020

To ascertain the characteristics of pool sloshing behavior that might be encountered during a core disruptive accident of sodium-cooled fast reactors, in our earlier work several series of experiments were conducted under various scenarios including the condition with mono-sized solid particles. It is found that under the particle-bed condition, three typical flow regimes (namely the bubble-impulsion dominant regime, the transitional regime and the bed-inertia dominant regime) could be identified and a flow-regime model (base model) has been even successfully established to estimate the regime transition. In this study, aimed to further understand this behavior at more realistic particle-bed conditions, a series of simulated experiments is newly carried out using mixed-size particles. Through analyses, it is verified that for present scenario, by applying the area mean diameter, our previously-developed base model can provide the most appropriate predictive results among the various effective diameters. To predict the regime transition with a form of extension scheme, a correction factor which is based on the volume-mean diameter and the degree of convergence in particle-size distribution is suggested and validated. The conducted analyses in this work also indicate that under certain conditions, the potential separation between different particle components might exist during the sloshing process.

• The Korean Journal of Pain
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• 제29권4호
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• pp.274-276
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• 2016

Chronic pelvic pain in women is a very annoying condition that is responsible for substantial suffering and medical expense. But dealing with this pain can be tough, because there are numerous possible causes for the pelvic pain such as urologic, gynecologic, gastrointestinal, neurologic, or musculoskeletal problems. Of these, musculoskeletal problem may be a primary cause of chronic pelvic pain in patients with a preceding trauma to the low back, pelvis, or lower extremities. Here, we report the case of a 54-year-old female patient with severe chronic pelvic pain after a transcutaneous electrical nerve stimulation (TENS) accident that was successfully managed with image-guided trigger point injections on several pelvic stabilizing muscles.

• 한국공작기계학회:학술대회논문집
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• 한국공작기계학회 2001년도 춘계학술대회 논문집(한국공작기계학회)
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• pp.34-39
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• 2001

The median barrier is one of the roadside hardware to prevent severe human and property damage from highway traffic accidents. The foreign standard of concrete median barrier was introduced and implemented without modification fitting to domestic vehicle and highway condition. In a car accident, median barrier doesn't protect vehicle effectively, especially for heavy vehicle such as bus and heavy truck. The purpose of this study is to develop the optimal performance design of concrete median barrier using the design of experiment with crash simulation analysis which is done by Pam-Crash that is one of the commercial crash simulation software. As a result of this study, an optimal design of concrete median barrier is obtained considering von Mises stress, volume and COG acceleration of truck.