• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2011년도 정기 학술대회
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• pp.400-403
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• 2011

본 논문에서는 시차열중량분석법과 X-선 회절분석법을 이용한 원전콘크리트의 탄산화에 의한 열화도 평가를 진행하였으며 두 가지 정성적 분석방법을 이용한 반정량적 평가 방법을 개발하였다. 원자력발전소 건설에 사용된 동일한 콘크리트 배합을 사용한 시편을 촉진 탄산화 시험장치에 28, 56, 91, 180, 365일 기간에 걸쳐 노출시켜 탄산화를 진행하였으며 노출된 시편은 시차열중량분석법, X-선 회절분석법을 이용하여 탄산화에 따라 발생된 열화생성물의 양을 정성적으로 분석하였다. 그 결과, 탄산화로 인해 발생되는 Calcite의 양이 노출기간에 따라 점차적으로 증가되는 것이 확인되었으며, Calcite의 생성을 위해 이산화탄소와 반응하는 Portlandite의 양이 점차적으로 감소되는 것이 확인되었다. 본 논문에서는 위의 언급된 두 방법의 관계성을 통해 열화도 평가를 진행하였다.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2011년도 정기 학술대회
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• pp.404-405
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• 2011

원전구조물의 방사성 폐기물 처분시설의 경우 지하수에 해수가 유입되어 콘크리트에 염소이온 침투가 발생할 수 있으며, 콘크리트 내부에 존재하는 인장철근의 부식에 의한 내구성 저하 및 수명 단축이 주된 문제가 된다. 본 논문에서는 원전콘크리트 구조물에서의 확산에 의한 염소이온 침투에 대한 수학적 모델을 제시하였다. 콘크리트 중의 염소이온의 침투는 콘크리트의 노출환경, 습윤상태에 따라 확산(Diffusion), 대류(Absorption), 전기적 이동(Migration)에 의해 발생한다. 이러한 조건을 모두 고려하여 제시한 방정식에 의해 염소이온의 침투를 예측할 수 있다.

• Journal of Radiation Protection and Research
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• 제47권2호
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• pp.99-106
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• 2022

Background: The radionuclide inventory calculation codes such as ORIGEN and FISPACT collapse neutron reaction libraries with energy spectra and generate an effective one-group cross-section. Since the nuclear cross-section data, energy group (g) structure, and other input details used by the two codes are different, there may be differences in each code's activation inventory calculation results. In this study, the calculation results of neutron-induced activation inventory using ORIGEN and FISPACT were compared and analyzed regarding radioactive waste classification and worker exposure during nuclear decommissioning. Materials and Methods: Two neutron spectra were used to obtain the comparison results: Watt fission spectrum and thermalized energy spectrum. The effective one-group cross-sections were generated for each type of energy group structure provided in ORIGEN and FISPACT. Then, the effective one-group cross-sections were analyzed by focusing on ⁵⁹Ni, ⁶³Ni, ⁹⁴Nb, ⁶⁰Co, ¹⁵²Eu, and ¹⁵⁴Eu, which are the main radionuclides of stainless steel, carbon steel, zircalloy, and concrete for decommissioning nuclear power plant (NPP). Results and Discussion: As a result of the analysis, ¹⁵⁴Eu and ⁵⁹Ni may be overestimated or underestimated depending on the code selection by up to 30%, because the cross-section library used for each code is different. When ORIGEN-44g, -49g, and -238g structures are selected, the differences of the calculation results of effective one-group cross-section according to group structure selection were less than 1% for the six nuclides applied in this study, and when FISPACT-69g, -172g, and -315g were applied, the difference was less than 1%, too. Conclusion: ORIGEN and FISPACT codes can be applied to activation calculations with their own built-in energy group structures for decommissioning NPP. Since the differences in calculation results may occur depending on the selection of codes and energy group structures, it is appropriate to properly select the energy group structure according to the accuracy required in the calculation and the characteristics of the problem.

• Earthquakes and Structures
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• 제17권5호
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• pp.511-519
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• 2019

The concrete gravity dam is one of the most important parts of the nation's infrastructure. Besides the benefits, the dam also has some potentially catastrophic disasters related to the life of citizens directly. During the lifetime of service, some degradations in a dam may occur as consequences of operating conditions, environmental aspects and deterioration in materials from natural causes, especially from dynamic loads. Cumulative Absolute Velocity (CAV) plays a key role to assess the operational condition of a structure under seismic hazard. In previous researches, CAV is normally used in Nuclear Power Plant (NPP) fields, but there are no particular criteria or studies that have been made on dam structure. This paper presents a method to calculate the limitation of CAV for the Bohyeonsan Dam in Korea, where the critical Peak Ground Acceleration (PGA) is estimated from twelve sets of selected earthquakes based on High Confidence of Low Probability of Failure (HCLPF). HCLPF point denotes 5% damage probability with 95% confidence level in the fragility curve, and the corresponding PGA expresses the crucial acceleration of this dam. For determining the status of the dam, a 2D finite element model is simulated by ABAQUS. At first, the dam's parameters are optimized by the Minitab tool using the method of Central Composite Design (CCD) for increasing model reliability. Then the Response Surface Methodology (RSM) is used for updating the model and the optimization is implemented from the selected model parameters. Finally, the recorded response of the concrete gravity dam is compared against the results obtained from solving the numerical model for identifying the physical condition of the structure.

• 콘크리트학회논문집
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• 제26권3호
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• pp.307-313
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• 2014

섬유보강 콘크리트의 균열 후 인장 거동을 예측하기 위해서는 균열면에 걸쳐 있는 섬유의 개수를 산정하는 섬유분포계수를 합리적으로 예측하는 것이 필요하다. 이 논문에서는 원형단면에서의 섬유분포계수를 분석하기 위해, 콘크리트 압축강도, 단면 크기, 섬유 종류 및 섬유혼입률 등을 변수로 강섬유보강 콘크리트 공시체를 제작하였으며, 제작한 공시체들을 타설 방향에 수직인 방향으로 절단한 후, 절단된 원형 단면에서의 섬유 개수로부터 섬유분포계수를 측정하였다. 측정 결과, 섬유가 타설면에 평행하게 분포할 확률이 증가함에 따라 실제 원형단면에서의 섬유분포계수가 일반적으로 알려진 0.5보다 작은 것으로 나타났다. 또한, 단위 면적 당 섬유 개수가 증가할수록 섬유분포계수가 감소하는 것으로 나타났다. 이 논문에서는 원형단면에서의 섬유분포계수를 합리적으로 예측하기 위해 섬유가 분포할 수 있는 각을 기하학적으로 분석하고, 이로부터 상세 모델과 단순화한 식을 유도하였다. 제안된 모델과 실험에서 측정된 섬유분포계수를 비교한 결과, 제안된 모델이 실제 원형단면에서의 섬유분포계수를 잘 예측하는 것으로 나타났다. 이 연구로부터 확보된 실험 결과 및 제안 모델은 향후 원형단면을 지닌 섬유보강 콘크리트 기둥 부재 등의 구조적 거동 연구에 매우 유용할 것으로 사료된다.

• 한국건설순환자원학회논문집
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• 제8권2호
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• pp.167-174
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• 2020

순환 골재 활용에 대한 연구가 국내에서도 비교적 많이 이루어져 왔으나, 대부분 순환 골재의 출처가 명확하지 않아, 국내 원자력 발전소와 같이 출처가 분명한 순환 골재를 재활용하는데 직접적으로 연구 결과를 적용하기에는 많은 불확실성이 존재한다. 따라서, 이 연구에서는 국내 원자력발전소 해체 시 발생하는 콘크리트 폐기물로부터 순환 굵은 골재를 생산 및 재활용할 수 있는 가능성에 대해 분석하기 위해, 국내 원자력발전소 모의 콘크리트를 제작 후 순환 굵은 골재를 생산하였다. 생산된 순환 굵은 골재를 활용하여 순환 굵은 골재 혼입률을 고려한 콘크리트를 배합하고, 역학적 특성을 실험적으로 분석하였다. 실험 결과 순환 굵은 골재 혼입률이 증가할수록 콘크리트 압축강도, 인장강도, 탄성계수 모두 전반적으로 감소하는 것으로 나타났으며, 순환 굵은 골재만을 사용한 경우 일반 콘크리트 대비 각각 최대 36, 37, 27% 정도로 감소하는 것으로 나타났다. 따라서, 향후 원자력발전소 해체로부터 생산된 순환 굵은 골재를 활용할 경우 혼입률에 대한 제한이 필요할 것으로 판단된다.

• Nuclear Engineering and Technology
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• 제53권5호
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• pp.1686-1704
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• 2021

This study addresses the numerical simulation of the shield building of an AP1000 nuclear power plant (NPP) subjected to a large commercial aircraft impact. First, a simplified finite element model (F.E. model) of the large commercial Boeing 737 MAX 8 aircraft is established. The F.E. model of the AP1000 shield building is constructed, which is a reasonably simplified reinforced concrete structure. The effectiveness of both F.E. models is verified by the classical Riera method and the impact test of a 1/7.5 scaled GE-J79 engine model. Then, based on the verified F.E. models, the entire impact process of the aircraft on the shield building is simulated by the missile-target interaction method (coupled method) and by the ANSYS/LS-DYNA software, which is at different initial impact velocities and impact heights. Finally, the laws and characteristics of the aircraft impact force, residual velocity, kinetic energy, concrete damage, axial reinforcement stress, and perforated size are analyzed in detail. The results show that all of them increase with the addition to the initial impact velocity. The first four are not very sensitive to the impact height. The engine impact mainly contributes to the peak impact force, and the peak impact force is six times higher than that in the first stage. With increasing initial impact velocity, the maximum aircraft impact force rises linearly. The range of the tension and pressure of the reinforcement axial stress changes with the impact height. The perforated size increases with increasing impact height. The radial perforation area is almost insensitive to the initial impact velocity and impact height. The research of this study can provide help for engineers in designing AP1000 shield buildings.

• Earthquakes and Structures
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• 제8권3호
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• pp.779-799
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• 2015

The slosh height and the possibility of water spill from rectangular Spent Fuel Storage Bays (SFSB) and Tray Loading Bays (TLB) of Nuclear power plant (NPP) are studied during 0.2 g, Safe Shutdown Earthquake (SSE) level of earthquake. The slosh height obtained through Computational Fluid dynamics (CFD) is compared the values given by TID-7024 (Housner 1963) and American concrete institute (ACI) seismic codes. An equivalent amplitude method is used to compute the slosh height through CFD. Numerically computed slosh height for first mode of vibration is found to be in agreement the codal values. The combined effect in longitudinal and lateral directions are studied separately, and found that the slosh height is increased by 24.3% and 38.9% along length and width directions respectively. There is no liquid spillage under SSE level of earthquake data in SFSB and TLB at convective level and at free surface acceleration data. Since seismic design codes do not have guidelines for combined excitations and effect of higher modes for irregular geometries, this CFD procedure can be opted for any geometries to study effect of higher modes and combined three directional excitations.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.2960-2973
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• 2022

Since Fukushima nuclear power plant (NPP) accident in 2011, the importance of research on various severe accident phenomena has been emphasized. Particularly, detailed analysis of combustion risk is necessary following the containment damage caused by combustion in the Fukushima accident. Many studies have been conducted to evaluate the risk of local hydrogen concentration increases and flame propagation using computational code. In particular, the potential for combustion by local hydrogen concentration in specific areas within the containment has been emphasized. In this study, the process of flame propagation generated inside a reactor core to containment during a loss of coolant accident (LOCA) was analyzed using MELCOR 2.1 code. Later in the LOCA scenario, it was expected that hydrogen combustion occurred inside the reactor core owing to oxygen inflow through the cold leg break area. The main driving force of the oxygen intrusion is the elevated containment pressure due to the molten corium-concrete interaction. The thermal and mechanical loads caused by the flame threaten the integrity of the containment. Additionally, the containment spray system effectiveness in this situation was evaluated because changes in pressure gradient and concentrations of flammable gases greatly affect the overall behavior of ignition and subsequent containment integrity.

• Nuclear Engineering and Technology
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• 제54권10호
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• pp.3849-3854
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• 2022

The present work aims to optimize the radiation protection efficiency for ion-selective containers used in the liquid treatment for the nuclear power plant (NPP) cooling cycle. Some naturally occurring rocks were examined as filler materials to reduce absorbed dose and equivalent dos received from the radioactive waste container. Thus, the absorbed dose and equivalent dose were simulated at a distance of 1 m from the surface of the radioactive waste container using the Monte Carlo simulation. Both absorbed dose and equivalent dose rate are reduced by raising the filler thickness. The total absorbed dose is reduced from 7.66E-20 to 1.03E-20 Gy, and the equivalent dose is rate reduced from 183.81 to 24.63 µSv/h, raising the filler thickness between 0 and 17 cm, respectively. Also, the filler type significantly affects the equivalent dose rate, where the redorded equivalent dose rates are 24.63, 24.08, 27.63, 33.80, and 36.08 µSv/h for natural rocks basalt-1, basalt-2, basalt-sill, limestone, and rhyolite, respectively. The mentioned results show that the natural rocks, especially a thicker thickness (i.e., 17 cm thickness) of natural rocks basalt-1 and basalt-2, significantly reduce the gamma emissions from the radioactive wastes inside the modified container. Moreover, using an outer cementation concrete wall of 15 cm causes an additional decrease in the equivalent dose rate received from the container where the equivalent dose rate dropped to 6.63 µSv/h.