• Nuclear Engineering and Technology
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• 제51권3호
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• pp.884-893
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• 2019

Spent fuel racks are steel structures used in the storage of the spent fuel removed from the nuclear power reactor. Rack units are submerged in the depths of the spent fuel pool to keep the fuel cool. Their free-standing design isolates their bases from the pool floor reducing structural stresses in case of seismic event. However, these singular features complicate their seismic analysis which involves a transient dynamic response with geometrical nonlinearities and fluid-structure interactions. An accurate estimation of the response is essential to achieve a safe pool layout and a reliable structural design. An analysis methodology based on the hydrodynamic mass concept and implicit integration algorithms was developed ad-hoc, but some dispersion of results still remains. In order to validate the analysis methodology, vibration tests are carried out on a reduced scale mock-up of a 2-rack system. The two rack mockups are submerged in free-standing conditions inside a rigid pool tank loaded with fake fuel assemblies and subjected to accelerations on a unidirectional shaking table. This article compares the experimental data with the numerical outputs of a finite element model built in ANSYS Mechanical. The in-phase motion of both units is highlighted and the water coupling effect is detailed. Results show a good agreement validating the methodology.

• 한국전산유체공학회지
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• 제21권2호
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• pp.47-53
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• 2016

This study investigated natural convection flow behavior in airflow path designed in concrete dry storage cask to remove the decay heat from spent nuclear fuels. Using FLUENT 16.1 code, thermal analysis for natural convection was carried out for three dimensional, 1/4 symmetry model under the normal condition that inlet ducts are 100% open. The maximum temperatures on other components except the fuel regions were satisfied with allowable values suggested in nuclear regulation-1536. From velocity and temperature distributions along the flow direction, the flow behavior in horizontal duct of air inlet and outlet duct, annular flow-path and bent pipe was delineated in detail. Theses results will be used as the theoretical background for the composing of airflow path for the designing of passive heat removal system by understanding the flow phenomena in airflow path.

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.624-634
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• 2016

Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that $k_{eff}$ values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.33-38
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• 1997

Criticality experiments with fixed neutron poison plates for water moderated and reflected low enriched(2.35 and 4.31 wt%) UO$_2$fuel rod clusters were evaluated to validate calculation techniques employed in analyzing fuel shipping and storage systems having steel, boral, or cadmium shield. Measurements were obtained for both the 2.35 wt% and the 4.31 wt% enriched rods in square pitched, water flooded lattices. The critical experiments with the 2.35 wt% enriched rods consists of three 20$\chi$ 16 or 20$\chi$ 17 fuel cluster. Critical separation were used in the experiments with the 4.31 wt% enriched fuel rods. In the experiments, the poison plates were placed on both sides of the centrally located fuel cluster. Critical separation between the three sub-critical fuel clusters were then measured for varying plate thicknesses and distances of the plates to the center fuel cluster. Calculations were performed for thirty eight critical configuration using KENO-V. a and MCNP. All of the results were within 1.23% in $\Delta$k when individually compared with the critical value of 1.0. Discrepancies of the code results are probably due to uncertainties in experiments and/or analytical modeling experiments. In general, MCNP predictions were observed to be in best agreement with the experiments.

• Nuclear Engineering and Technology
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• 제51권4호
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• pp.1117-1131
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• 2019

Tanks are used extensively in many engineering areas for spent fuel pool structures at nuclear power plants or for water storage tanks in bulk carriers. To ensure the structural integrity of such tanks when under dynamic loads, modal characteristics such as natural frequencies, participation factors and mode shapes should be known. Investigated in this study are the modal characteristics of a square tank by the finite element method. This approach can be used with subsequent dynamic analyses such as a response spectrum analysis or a harmonic analysis. Finite element models are prepared to determine the natural frequencies and mode shapes, which are easy to find the modal characteristics of a fluid-filled square tank. The effects of the fluid contained in the tank and the boundary conditions at top and bottom ends on the modal characteristics are assessed by several finite element analyses.

• Journal of Radiation Protection and Research
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• 제40권2호
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• pp.92-100
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• 2015

경수로 사용후핵연료 건식 중간저장시설의 방사선영향평가 효율성 개선을 목적으로 '선원항 지정방법에 따른 민감도 평가', '2-Step 계산'기법 개발 및 '냉각기간 이득효과' 적용에 따른 방사선 영향평가를 수행하였다. 본 연구에서는 저장건물의 용기배열 순서에 따라 순차적으로 선원항을 지정하여 직접선량에 미치는 민감도를 평가하였으며, 차폐건물 외벽에서의 방사선량은 내벽과 인접한 최근접 2개 열에 의한 영향이 지배적임을 확인하였다. 또한, 저장시설에 차폐 건물이 도입될 경우, 막대한 전산해석 시간을 감소시키기 위해 '2-Step 계산'기법을 수립하여 평가한 결과는 절반가량의 해석시간으로 직접(1-Step) 계산결과와 유사한 결과를 도출하였다. 마지막으로, 저장시설에 순차적으로 저장되는 저장용기의 보관기간을 사용후핵연료의 실제 냉각기간을 적용하면 건물 외벽에서의 방사선량이 냉각기간을 모두 동일하게 설정한 계산값에 비해 40% 정도 낮게 평가됨을 확인하였다. 본 연구는 중간저장시설의 방사선 영향평가를 위한 몬테칼로 차폐해석 방법의 효율성을 향상시키고자 수행되었으며, 좀 더 다양한 사례에 대한 평가를 통하여 신뢰성을 향상시킨다면 저장시설의 설계 및 부지경계 기준설정에 활용할 수 있을 것이다.

• 방사선산업학회지
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• 제18권1호
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• pp.35-42
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• 2024

Demand for RW transportation is expected to increase due to the continuous generation of RW from nuclear power plants and facilities, decommissioning of plants, and saturation of spent fuel temporary storage facilities. The locational aspect of plants and radiation protection optimization for the public have led to an increasing demand for maritime transportation, necessitating to apprehend the overseas and domestic current status. Given the potential long-term radiological impact on the public in the event of a sinking accident, a pre-transportation exposure assessment is necessary. The objective of this study is to investigate the overseas and domestic RW maritime transportation current status and overseas dose assessment cases for the public in sinking accident. Selected countries, including Japan, UK, Sweden, and Korea, were examined for transport cases, Japan and the U.S were chosen for dose assessment case in sinking accidents. As a result of the maritime transportation case analysis, it was performed between nuclear power plants and reprocessing facilities, from plants to disposal or intermediate storage facilities. HLW and MOX fuel were transported using INF 3 shipments, and all transports were performed low speed of 13 kn or less. As a result of the dose assessment for the public in sinking accident, japan conducted an assessment for the sinking of spent fuel and vitrified HLW, and the U.S conducted for the sinking of spent fuel. Both countries considered external exposure through swimming and working at seashore, and internal exposure through seafood ingestion as exposure pathway. Additionally, Japan considered external exposure through working on board and fishing, and the U.S considered internal exposure through spray inhalation and desalinized water and salt ingestion. Internal exposure through seafood ingestion had the largest dose contribution. The average public exposure dose was 20 years after the sinking, 0.04 mSv yr^-1 for spent fuel and 5 years after the sinking, 0.03 mSv yr^-1 for vitrified HLW in Japan. In the U.S, it was 1.81 mSv yr^-1 5 years after the sinking of spent fuel. The results of this study will be used as fundamental data for maritime transportation of domestic RW in the future.

• 방사성폐기물학회지
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• 제8권3호
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• pp.239-245
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• 2010

2009년말 기준으로 11,811 다발의 경수로 사용후핵연료가 방출되었으며, 지금까지 각 사용후핵연료에 대해 방사선원항을 가중하여 설계에 반영하기는 사실상 불가능하여, 원자력 관련시설 설계시 보수성을 갖는 기준 사용후핵연료를 선정하고 이를 바탕으로 시스템 설계를 수행하여 왔다. 방사선원항에 대한 단순모델을 적용하면 각 사용후핵연료에 대한 방사선원항을 가중함으로써 이와 같은 보수성을 배제할 수 있으므로 본 연구에서 웨스팅하우스형 원전에 사용된 사용후핵연료를 대상으로 방사선원항, 즉, 붕괴열, 방사능세기, 섭취위해도 등을 예측하기 위한 회귀모형을 개발하였다. 개발된 회귀식을 통해 예측된 방사선원항값은 ORIGEN-ARP 코드로 계산된 값과 약 5% 이내에서 잘 일치함을 확인하였으며, 이의 유용성을 검토한 결과 각각의 사용후핵연료에 대한 방사선원항을 가중하여 설계에 반영하면 보수성을 줄일 수 있음을 확인하였다. 따라서 본 연구에서 개발된 회귀식은 사용후핵연료의 저장 및 처분과 관련한 원자력시설 설계시 개념설계 단계에서 유용하게 사용될 수 있을 것으로 판단된다.

• 방사선산업학회지
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• 제18권1호
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• pp.9-14
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• 2024

In order to thoroughly verify the denuclearization of the Korean Peninsula, it is urgent to develop technology capabilities to monitor, detect, collect, analyze, interpret, and evaluate nuclear activities using nuclear materials and secure nuclear transparency. The IAEA is actively using seal technology to maximize the efficiency of safety measures, and currently uses metal cap, paper, COBRA, and EOSS as seal devices. Unlike facilities that comply with safety measures requirements, such as domestic nuclear facilities, facilities subject to denuclearization are likely to have various risk environments that make it difficult to apply safety measures, and there is a high possibility that continuity of knowledge (COK) such as damage, malfunction, and power loss will not be maintained. This study aims to develop a real-time active seal device that can be applied in such special situations to enable immediate response in the event of a similar situation. To this end, the main functions of the real-time seal device were derived and applied, and a commercialized seal device and operation software. The real-time seal technology developed through this study can be applied to all nuclear facilities in South Korea, especially used as storage equipment for dry cask storage facilities of heavy water reactor's after fuel, and it is believed that unnecessary radiation exposure by inspectors can be minimized.

• 방사성폐기물학회지
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• 제20권1호
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• pp.127-131
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• 2022

The structural safety of prototype transport and storage containers for very low-level radioactive liquid waste was experimentally estimated for its localization development. Transport containers for radioactive liquid waste have been researched and developed, however, there are no standardized commercial containers for very low-level radioactive waste in Korea. In this study, the structural safety of the designated IP-2 type container capable of transporting and temporarily storing large amounts of very low-level liquid waste, which is generated during the operation and decommissioning of nuclear power plants, was demonstrated. The stacking and drop tests, which were conducted to determine the structural integrity of the container, verified that there was no external leakage of the contents in spite of its structural deformation due to the drop impact. This study shows the effort required for the localization of the technology used in manufacturing transport and storage containers for very low-level radioactive liquid waste, and the additional structural reinforcement of the container in which the commercial intermediate bulk container (IBC) external frame was coupled.