Composite laminates are used in a wide range of applications including defense, automotive, aviation and aerospace, marine, wind energy, and recreational sporting goods. These composite beams still exhibit problems such as buckling, local deformations, and interlaminar delamination. To overcome these drawbacks, a novel viscoelastic autoclave bonding with tapered epoxy reinforcement polyurethane films is proposed. In existing laminates, compression face wrinkling and interlaminar delamination is caused in the sandwich beam. The unique viscoelastic autoclave spunbond interlayer bonding is designed to prevent face wrinkling and absorb and distribute stresses induced by external loads, thereby eliminating interlaminar delamination in the sandwich beam. Also, the existing special reinforcement causes stress concentrations, and the core is not effectively connected, which directly affects the stiffness of the beam. To address this, a novel tapered epoxy polyurethane reinforcement adhesive film is proposed, whose reinforcement thickness gradually tapers as it enters the core material. This minimizes stress concentrations at the interface, preventing excessive adhesive squeeze-out during the bonding process, and improves the stiffness of the beam. Results indicate the proposed model avoids the formation of micro cracks, interlaminar delamination, buckling, and local deformations, and effectively improves the stiffness of the beam.

The operation time of a disposal repository is generally more than one hundred years except for the institutional control phase. The structural integrity of a repository can be regarded as one of the most important research issues from the perspective of a long-term performance assessment, which is closely related to the public acceptance with regard to the nuclear safety. The objective of this study is to suggest the methodology for quantitative evaluation of structural integrity in a nuclear waste repository based on the adaptive artificial intelligence (AI), fractal theory, and acoustic emission (AE) monitoring. Here, adaptive AI means that the advanced AI model trained additionally based on the expert’s decision, engineering & field scale tests, numerical studies etc. in addition to the lab. test. In the process of a methodology development, AE source location, wave attenuation, the maximum AE energy and crack type classification were subsequently studied from the various lab. tests and Mazars damage model. The developed methodology for structural integrity was also applied to engineering scale concrete block (1.3 m × 1.3 m × 1.3 m) by artificial crack generation using a plate jacking method (up to 30 MPa) in KURT (KAERI Underground Research Tunnel). The concrete recipe used in engineering scale test was same as that of Gyeongju low & intermediate level waste repository. From this study, the reliability for AE crack source location, crack type classification, and damage assessment increased and all the processes for the technology development were verified from the Korea Testing Laboratory (KTL) in 2022.

A lot of CANDU Spent Fuels (CSFs) have been stored in spent nuclear fuel pools and dry storage facilities. In accordance with the enhanced nuclear regulations, the initial characteristics of CSF should be inspected to ensure the integrity of CSF and the reliable operation of storage system before loading it into a cask for long-term dry storage. For the inspections, an initial characteristics measurement equipment was designed, which is used for Pool-Side Examination (PSE) in the spent fuel pool of the pressurized heavy water reactor nuclear power plant. Measurements using the equipment consist of non-contact inspections and contact inspections. The non-contact inspections do not affect CSF integrity, whereas the integrity of CSF can be reduced during the contact inspections under abnormal operating conditions because the probe of equipment may apply specific loads to the CSF. Therefore, the structural integrity evaluations of equipment and CSF are performed using Finite Element (FE) analyses for four combinations based on two abnormal conditions and two probe positions. The used abnormal conditions are the pressing load condition and the scratching load condition, and two probe positions are the center and bottom of the fuel rod in the longitudinal direction, respectively. In this evaluation, the bottoms of the fuel rod or CSF are defined as the regions facing the bottom surface of equipment. The analysis of the pressing load condition is performed by pressing the probe of the equipment in radial direction of the CSF fuel rod. That of the scratching load condition is carried out by applying a specific radial load to the CSF fuel rod using the probe and then applying the load to the surface of the fuel rod while moving axially along the surface. All combinations are analyzed considering geometric, boundary and material non-linearity under the dynamic load, which is dependent on the equipment operating velocity. The stresses of CSF and equipment components were obtained from these analyses. The maximum stress of each component was generated at the combination on the scratching load condition for the bottom position among the four combinations. The obtained maximum stresses are lower than the yield stress for each component material. Also, the CSF is not overturned due to the support plate of the equipment in all analyses. Therefore, the structural integrity and safety of the equipment and the CSF are maintained under abnormal operating conditions during the inspection using the initial characteristic measurement equipment.

On-site storage facility using concrete silo dry storage systems for spent nuclear fuel at Wolsong NPP site came into operation in 1992 and was expanded four times, and a total of 300 silo dry storage systems are currently in operation. The design lifetime of silo dry storage systems has been licensed for 50 years. As the dry storage systems are subject to time constraints for a limited lifetime, countries operating the dry storage systems are working to ensure the long-term integrity of dry storage systems and IAEA also recommends that the dry storage systems be assessed for long-term storage. To demonstrate the long-term integrity due to material degradation during the licensed design lifetime, the structural integrity of silo dry storage systems was evaluated by considering the material degradation characteristics of concrete. The concrete compressive strength results measured so far by the rebound hammer method, which is an internationally standardized nondestructive test method for converting hardness into compressive strength using the correlation between rebound number and strength at the time of a Schmidt hammer strike, were analyzed in accordance with Wolsong NPP’s procedure to quantify the degradation characteristics, and the prediction of concrete strengths for 20 years and 50 years after construction of the silo dry storage systems was determined, respectively. Based on these residual compressive strengths, structural analyses of the silo dry storage systems were carried out under normal, off-normal and accident conditions of the related regulations, and the structural integrity of silo dry storage systems was reevaluated. It was confirmed the silo dry storage systems are able to maintain structural integrity up to the design lifetime of 50 years even if the concrete is deteriorated.

The Deep Borehole Disposal (DBD) method has various advantages, such as minimizing the use of site area and corrosion of the disposal container and improving long-term structural safety. However, it is necessary to review the problems that may occur in various technologies related to the emplacement and retrieval of the disposal container and the sealing of the borehole. Therefore, the purpose of this study is to evaluate the structural integrity of an emplacement and retrieval device (hereinafter, the disposal container connecting device) of a DBD container. The disposal connecting device was evaluated according to ANSI 14.6 and NUREG-0612 standards. The allowable stress should be less than the yield strength under the load condition of 3g. The length of the disposal container connecting device was about 2,900 mm, the diameter was 406 mm, and the weight was about 1.2 tons. In addition, 10 disposal containers weighing up to 2.2 tons were handled. The disposal container connecting device was made of stainless steel, and the maximum operating temperature was about 300°C. For structural evaluation, ABAQUS finite element analysis program was used. The analysis model was modeled only 1/2 part considering symmetry condition. The analysis model was modeled using 410,431 nodes and 344,119 solid elements. Three times load was applied to the weight of the disposal container. Axisymmetric conditions were applied to the symmetrical surface of the disposal container, and vertical restraints were applied to the upper lifting lugs. A surface-to-surface contact condition was applied to the part where the contact occurred. As a result of the analysis, the greatest stress was generated at the part supported by the clamp at the disposal container connector at 168.9 MPa. In the lugs and pins connecting the guide and the connecting device, a stress of 530.1 MPa was generated by shearing. In the bolts of the disposal container connecting device, a stress of 498MPa was generated and the safety margin was 1.73. A stress of 486.1 MPa was generated in the disposal container connecting device, and the safety margin was the smallest 1.16. As a result of the analysis, all components of the disposal container connecting device showed a safety margin of 1.16 or more at the maximum operating temperature and satisfied the allowable stress.

With respect to spent nuclear fuels, disposal containers and bentonite buffer blocks in deep geological disposal systems are the primary engineered barrier elements that are required to isolate radioactive toxicity for a long period of time and delay the leakage of radio nuclides such that they do not affect human and natural environments. Therefore, the thermal stability of the bentonite buffer and structural integrity of the disposal container are essential factors for maintaining the safety of a deep geological disposal system. The most important requirement in the design of such a system involves ensuring that the temperature of the buffer does not exceed 100℃ because of the decay heat emitted from high-level wastes loaded in the disposal container. In addition, the disposal containers should maintain structural integrity under loads, such as hydraulic pressure, at an underground depth of 500 m and swelling pressure of the bentonite buffer. In this study, we analyzed the thermal stability and structural integrity in a deep geological disposal environment of the improved deep geological disposal systems for domestic light-water and heavy-water reactor types of spent nuclear fuels, which were considered to be subject to direct disposal. The results of the thermal stability and structural integrity assessments indicated that the improved disposal systems for each type of spent nuclear fuel satisfied the temperature limit requirement (< 100℃) of the disposal system, and the disposal containers were observed to maintain their integrity with a safety ratio of 2.0 or higher in the environment of deep disposal.

최근 국내에서 육상 및 해상을 통한 소외 정상운반 시 진동 및 충격하중에 대한 사용후핵연료의 건전성 평가 기술 개발이 수행되고 있다. 이와 관련된 국내 연구사례는 전무하여 기존에 진행된 또는 현재 수행중인 해외연구사례를 조사하여 국내 연구에 참고하고자 한다. 2000년 이전 과거 미국의 사용후핵연료의 정상운반 시 진동 및 충격하중 측정 관련 연구현황을 조사 하였고 2009년부터 미국국립연구소 주관으로 실시한 단축가진시험, 콘크리트블럭 트럭운반시험, 다축가진시험에 대해서 조사하였으며 2017년 미국 SNL, 스페인의 ENSA, 한국이 공동으로 수행한 복합운반시험을 상세히 조사하였다. 시험 준비과정, 절차, 가속도 및 변형률 측정결과, 유한요소 및 다물체동역학 해석과정 등이 조사되었다. 각 시험 별로 측정된 변형률 자료를 바탕으로 사용후핵연료 피로곡선과 비교한 결과 손상을 일으키기에는 매우 미미한 정도의 변형률이 발생한다는 초기 결론을 얻었음을 확인하였다. 하지만 현재 결론은 일부 결과만을 검토한 예비 결론으로 상세한 검토가 현재 미국에서 진행 중이다. 미국에서 지금까지 수행한 사용후핵연료의 정상운반조건에서의 진동 및 충격하중 측정과 관련하여 조사된 내용은, 국내 운반환경에서 사용후핵연료의 정상운반시험을 수행할 때 참고할만한 유용한 자료라 판단된다.

본 연구는 수중 및 여가활동에 대한 수요 증가에 따른 다이버들을 위한 보트의 구조 건전성에 관한 것이다. 대상 선박은 선체 중앙부에 Moon Pool 구조를 갖추고 있는 소형 쌍동선이며, 연구수행은 ISO Rule 기반의 허용응력 산정을 통한 유한요소 해석법을 이용하여 연구를 수행하였다. 연구수행 방법은 ISO 12215-5와 TC118.1225-7에서 정의하고 있는 계수를 산정하고, 종방향굽힘 모멘트, 비틀림 모멘트, 선저슬래밍 하중 등을 적용하여 ISO 기준과 허용응력 설계법(ASD)에 의한 적합성 여부를 판정하고 유한요소해석(FEA)를 활용한 극한강도 설계법을(LFRD)를 적용하여 수행하였다. 연구결과 문풀형 구조를 가진 선박도 ISO규정, KR규정을 적용하여 설계시 구조적 건전성을 확보 하는 것으로 사료된다.

가스 터빈은 기동 및 정지 횟수가 많기 때문에 열피로나 취화 현상으로 인한 가스터빈 케이싱의 균열 또는 케이싱의 플랜 지면에서 고온고압 가스의 누설이 발생할 가능성이 높다. 따라서 가스터빈 케이싱의 구조안전성 및 플랜지면에서의 누설평 가는 반드시 수행되어야 하는 부분이다. 본 논문에서는 유한요소해석을 바탕으로 터빈 케이싱의 ASME B&PVC VIII-2 구 조안전성 평가 및 접촉압력을 통한 누설 평가 그리고 볼트의 구조안전성 평가를 진행하였다. 또한 가스터빈 케이싱의 유한 요소모델링 및 해석/평가 방법을 제안하여 가스터빈 개발에 활용할 수 있게 하였다.

가스 터빈은 기동 및 정지 횟수가 많기 때문에 열피로나 취화 현상으로 인한 가스터빈 케이싱의 균열 또는 케이싱의 플랜지면에서 고온고압 가스의 누설이 발생할 가능성이 높다. 따라서 가스터빈 케이싱의 구조안전성 및 플랜지면에서의 누설평가는 반드시 수행되어야 하는 부분이다. 본 논문에서는 유한요소해석을 바탕으로 터빈 케이싱의 ASME B&PVC VIII-2 구조안전성 평가 및 접촉압력을 통한 누설 평가 그리고 볼트의 구조안전성 평가를 진행하였다. 또한 가스터빈 케이싱의 유한요소모델링 및 해석/평가 방법을 제안하여 가스터빈 개발에 활용할 수 있게 하였다.

In this paper, as the transport cask was moved in the reactor, the structural integrity on the cask had to be evaluated in the normal transport condition. The drop height of the cask was determined by the weight of the cask in the normal transport condition by regulations about assessment test. It was determined that the drop height of the cask was 1.2 m by regulations. The velocity of the drop impact was calculated to perform the drop impact analysis by the principle of the conservation of energy. Using results of the simulation about the drop impact analysis, the structural integrity assessment on the transport cask was performed by ASME Boiler and Pressure Vessel Code.

원자로 격납건물은 냉각재상실사고와 같이 내부의 과도한 압력이 유발되는 사고에 있어서도 방사성 물질이 외부로 누출되지 않도록 막는 최종의 방벽이다. 이러한 격납건물의 기능적 중요성에 기인하여, 건설 초기 구조건전성시험(SIT)을 수행한다. 이러한 SIT거동을 가장 실제와 가깝게 예측하기 위한 해석 연구를 수행하였다. 해당 연구의 결과는 2편의 논문으로 정리되었는데, 본 논문은 그 중 II편으로 I편의 해석모델 구성 시의 주요 고려사항의 분석 및 예비해석 결과를 반영한 상세 해석 모델의 구성 과정 및 해석 결과를 제시하고 있다. 특히 비부착식 텐던으로 시공된 구조물에서 덕트관에 의한 강성 저감효과 및 덕트관을 사이에 둔 텐던과 콘크리트간의 밀착 여부에 따른 영향을 해석 시 최대한 고려하고자 하였다. 이러한 과정을 통해 구축된 해석 모델에 따른 변위과 신고리 3호기 SIT 측정변위를 비교한 결과, ASME CC-6000 기준을 충분히 만족시키는 결과가 나타남을 확인하였다.

원자로 격납건물은 냉각재상실사고와 같이 내부의 과도한 압력이 유발되는 사고에 있어서도 방사성 물질이 외부로 누출되지 않도록 막는 최종의 방벽이다. 이러한 격납건물의 기능적 중요성에 기인하여, 건설 초기 구조건전성시험(SIT)을 수행한다. 신고리 3호기 SIT 시험 당시 계측된 변위를 예측하기 위한 초기 해석 모델은 일부 위치에서 실제 변위를 과소 평가하는 경향을 보임에 따라 이를 개선하고자 하는 연구가 수행되었다. 해당 연구의 결과를 I 편과 II 편의 논문으로 정리하였으며, 본 I 편에서는 초기 해석모델을 개선해가는 과정에서의 해석모델 구성 시의 주요 고려사항의 분석 및 예비해석 결과를 제시하고 있다. 우선적으로 콘크리트 자체의 해석요소(mesh) 구성과 라이너, 철근, 텐던 등의 요소간의 연결 설정이 중요함을 확인하였다. 또한, 다양한 예비해석의 결과를 통해 비부착식 텐던으로 시공된 구조물에서 덕트관에 의한 강성 저감 효과 및 덕트관을 사이에 둔 텐던과 콘크리트간의 밀착 여부에 따른 강성 영향을 적절히 고려하는 것이 중요함을 확인하였다.

As the world energy consumption grows, the interest in marin energy resources is increasing. In excavating such resources, the marine riser which connects the floating structure and sea bed is an essential device. The riser system is often exposed to harsh ocean environment and thus vulnerable to damage. Since the failure of the riser system may cause serious economical loss as well as environmental problem, the structural integrity of the riser is very important. Generally, the riser is an extremely slender structure with a much smaller diameter than a length. Therefore, a structural integrity monitoring methodology for typical buildings and bridges may not be applicable. In this paper, the applicability of a damage identification method for a structure to a marine riser is examined via a numerical example. Also, recent research practices and findings for monitoring the behavior and the structural integrity of the marine riser are examined and summarized.

본 논문에서는 노심용융사고 시 관통노즐이 제거된 원자로용기 하부헤드의 구조 건전성 평가를 수행하였다. 열응력, 노심용융물의 질량 그리고 내압조건의 해석결과를 고려할 때, 하부헤드의 열응력에 의한 영향이 가장 크게 나타났다. 손상 가능성은 파손기준에 따라 평가하였으며, 등가소성변형률이 임계변형률 파손기준보다 낮은 수준으로 평가되었다. 열-구조물 연성해석 결과 하부헤드의 두께 중간층에서 항복강도보다 낮은 응력이 발생한 탄성영역 구간을 확인하였다. 내압이 커지면서 탄성영역 범위가 점차 좁아지면서 탄성영역이 내벽으로 이동하는 결과를 확인하였고, 노심용융사고 시 구조적 건전성을 만족하는 것으로 평가되었다.

폴리머 시멘트 고화체는 일반 몰타르 내의 시멘트 수화물을 폴리머 개질제를 이용하여 부분적으로 대 체함으로써 그 기능을 강화시킨 복합재료로써, 특히 시멘트 몰타르에 폴리머를 첨가하는 것은 그 화학적 내구성을 향상시킨다고 알려져 있다. 따라서 본 연구에서는 고화재료로서의 폴리머 시멘트에 대한 낮은 침투성 및 낮은 이온 확산도 등과 같은 향상된 화학적 내구성을 확인하기 위하여 폴리머 시멘트 시편들을 제조하였다. 이때 폴리머의 함량은 0에서부터 30%까지 변화시켰으며, 물에 대한 시멘트 비(W/C)를 33%와 50%로 각각 유지 시켰다. 충분히 경화시킨 후에, 제조된 시편들에 대한 구조적 건전성을 압축강도와 수침법에 의한 공극도를 통하여 평가하였다. 그 결과, W/C 비가 33%이고, 폴리머 함량이 약 10%인 폴리 머 시멘트 시편에서 가장 향상된 개질변화를 얻을 수 있었다. 끝으로 이 최적의 조합비를 가지는 시편에 대하여 ANS 16.1에 따르는 침출시험을 수행하였으며, 그 결과를 일반 시멘트 고화체와 비교하였다.