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.

Integrity evaluation scheme for Spent Fuel (SF) dry storage has been developed under transportation failure modes. This method especially considered the degradation characteristics of Spent Fuel (SF) during dry storage such as radial and circumferential hydride content, hydride volume fraction, oxide thickness, etc. Hydride and zircaloy cladding are considered as material composite system, using correlation models related to material properties. Critical Strain Energy Density (CSED) is compared with Strain Energy Density (SED), to evaluate cladding integrity. CSED serves as material characteristics, while SED can be considered as boundary condition. To calculate the CSED of cladding in the lateral failure mode, circumferential hydride concentration is used. SED is calculated considering both the bending moment and axial load. On the other hand, in the longitudinal failure case, fuel rod temperature, internal pressure, hoop stress, radial hydride concentration is used to calculate CSED. And pinch force (contact) was considered to evaluate SED. Model validations were conducted by comparing hot cell SF test and existing validated evaluation results. To separately handle normal transportation conditions from hypothetical accident conditions, SED according to stress-strain analysis results was separated into elastic and plastic regions. As a result of applying this scheme for 14×14 SF, failure probability of normal condition was zero, which is the similar result with DOE and same with EPRI. Regarding accident condition, lateral case showed similar result, but longitudinal case showed different but reasonable result, which was due to the different analysis conditions. The proposed methodology which was indigenously developed through this study is named as K-method.

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.

The safety of a KTC-360 transport cask, a large-capacity pressurized heavy-water reactor transport cask that transports CANDU spent nuclear fuel discharged from the reactor after burning in a pressurized heavy-water reactor, must be demonstrated under the normal transport and accident conditions specified under transport cask regulations. To confirm the thermal integrity of this cask under normal transport and accident conditions, high-temperature and fire tests were performed using a one-third slice model of an actual KTC-360 cask. The results revealed that the surface temperature of the cask was 62°C, indicating that such casks must be transported separately. The highest temperature of the CANDU spent nuclear fuel was predicted to be lower than the melting temperature of Zircaloy-4, which was the sheath material used. Therefore, if normal operating conditions are applied, the thermal integrity of a KTC-360 cask can be maintained under normal transport conditions. The fire test revealed that the maximum temperatures of the structural materials, stainless steel, and carbon steel were 446°C lower than the permitted maximum temperatures, proving the thermal integrity of the cask under fire accident conditions.

HIC refers to a radwaste packaging container that can maintain integrity for more than 300 years in the general underground environment and disposal conditions in Korea. For HIC, the integrity of containers is verified according to the HIC regulation guideline for LLW and ILW disposal. Existing material tests include mechanical stability, permeability resistance, corrosion properties, chemical durability and biological resistance. In this study, a chemical durability test was conducted to prove the suitability of the HIC material by measuring the degree of chemical influence other than corrosion from the disposal environment. The chemical resistance evaluation method was used to simulate the disposal environment in the underground repository, and the amount of change in the physical properties of the degraded polymer concrete specimens according to the test time was confirmed. The technical standards considered leaching of material components, sulfation attack, acid attack, alkali, carbonate, and salt crystallization. The compressive strength and weight change of the specimens with time were checked by immersing them in a chemical solution that could leak major hazardous substances and wastes in the groundwater of the repository for several months. In addition, in order to evaluate the integrity in condition severe than the disposal environment, a flow was applied to a chemical solution having a concentration twice that of the basic chemical resistance test conditions, and the test period was extended twice to accelerate the deterioration of the specimen.

Prior to the investigations on fuel degradation it is necessary to describe the reference characteristics of the spent fuel. It establishes the initial condition of the reference fuel bundle at the start of dry storage. In a few technology areas, CANDU fuels have not yet developed comprehensive analysis tools anywhere near the levels in the LWR industry. This requires significantly improved computer codes for CANDU fuel design. In KNF, in-house fuel performance code was developed to predict the overall behavior of a fuel rod under normal operating conditions. It includes the analysis modules to predict temperature, pellet cracking and deformation, clad stress and strain at the mid-plane of the pellet and pellet-pellet interfaces, fission gas release and internal gas pressure. The main focus of the code is to provide information on initial conditions prior to dry storage, such as fission gas inventory and its distribution within the fuel pellet, initial volumes of storage spaces and their locations, radial profile of heat generation within the pellet, etc. Potential degradation mechanisms that may affect sheath integrity of CANDU spent fuel during dry storage are: creep rupture under internal gas pressure, sheath oxidation in air environment, stress corrosion cracking, delayed hydride cracking, and sheath splitting due to UO2 oxidation for a defective fuel. To upgrade the developed code that address all the damage mechanisms, the first step was a review of the available technical information on phenomena relevant to fuel integrity. The second step was an examination of the technical bases of all modules of the in-house code, identify and extend the ranges of all modules to required operating ranges. Further improvements being considered include upgrades of the analysis module to achieve sufficient accuracy in key output parameters. The emphasis in the near future will be on validation of the in-house code according to a rigorous and formal methodology. The developed models provide a platform for research and industrial applications, including the design of fuel behavior experiments and prediction of safe operating margins for CANDU spent fuel.

This paper aims to evaluate the mechanical integrity for Spent Nuclear Fuel (SNF) cladding under lateral loads during transportation. The evaluation process requires a conservative consideration of the degradation conditions of SNF cladding, especially the hydride effect, which reduces the ductility of the cladding. The dynamic forces occurring during the drop event are pinch force, axial force and bending moment. Among those forces, axial force and bending moment can induce transverse tearing of cladding. Our assessment of 14 × 14 PWR SNF was performed using finite element analysis considering SNF characteristics. We also considered the probabilistic procedures with a Monte Carlo method and a reliability evaluation. The evaluation results revealed that there was no probability of damage under normal conditions, and that under accident conditions the probability was small for transverse failure mode.

This study developed an analytical methodology for the mechanical integrity of spent nuclear fuel (SNF) cladding tubes under external pinch loads during transportation, with reference to the failure mode specified in the relevant guidelines. Special consideration was given to the degraded characteristics of SNF during dry storage, including oxide and hydride contents and orientations. The developed framework reflected a composite cladding model of elastic and plastic analysis approaches and correlation equations related to the mechanical parameters. The established models were employed for modeling the finite elements by coding their physical behaviors. A mechanical integrity evaluation of 14 × 14 PWR SNF was performed using this system. To ensure that the damage criteria met the applicable legal requirements, stress-strain analysis results were separated into elastic and plastic regions with the concept of strain energy, considering both normal and hypothetical accident conditions. Probabilistic procedures using Monte Carlo simulations and reliability evaluations were included. The evaluation results showed no probability of damage under the normal conditions, whereas there were small but considerably low probabilities under accident conditions. These results indicate that the proposed approach is a reliable predictor of SNF mechanical integrity.

In this study, we examined number, motility and plasma membrane integrity of spermatozoa from six regions of epididymis in bull. Six testicles with epididymides were castrated from six bulls (mean±standard error, age of days = 441.3±9.6, body weight (kg) = 367±8.4, scrotal circumference (cm) = 30.7±0.4) at Hanwoo Research Institute, NIAS and transported to laboratory within 1 hour. Testicular weight, length, width and circumference were recorded. Epididymis in each bull was randomly used for recovery of spermatozoa. Epididymis was divided into six regions: efferent duct (ED), caput, corpus, proximal cauda (Pcauda), distal cauda (Dcauda) and vas deferens (VD). In experiment 1, we examined sperm number of each region of epididymis. Each region of epididymis contained different number of spermatozoa: ED (37.8±15.7 × 106cells/ml, 8.2%), caput (93.6±18.8 × 106cells/ml, 20.2%), corpus (33.0±8.5 × 106cells/ml, 7.1%), Pcauda (104.2±23.5 × 106cells/ml, 22.5%), Dcauda (180.5±32.5 × 106cells/ml, 39.0%) and VD (14.0±5.0 × 106cells/ml, 3.0%). In experiment 2, sperm motility of each epididymal region was examined by computer assisted sperm analysis (SCA, MicroOptic) system. Sperm motility was divided into 4 groups (fast progressive, slow progressive, non-progressive and immotile) based on WHO guideline. Percentages of fast progressive of Pcauda and Dcauda (11.0±2.3 and 15.4±3.6%) were significantly higher than that of ED, Caput, Corpus and VD which is 0.1±0.1, 1.5±0.6, 1.9±0.7 and 0.3±0.2%, respectively (p<0.05). In experiment 3, percentage of intact plasma membrane spermatozoa of each regions were examined by hypoosmotic swelling test. Percentages of intact plasma spermatozoa were not significantly different among six regions of epididymis: ED, caput, corpus, Pcauda, Dcauda and VD which is 68.0±8.6, 74.0±5.3, 68.5±6.2, 70.8±5.5, 71.0±5.8 and 64.6±10.8%, respectively. In conclusion, in the present study, we found out distribution, motility and plasma membrane integrity of spermatozoa from six regions of epididymis in Hanwoo bull. These results will be contributed to basic research about spermatozoa transportation and characters in epididymis of bull.

In this study, we examined total number, motility and plasma membrane integrity of epididymal spermatozoa from cauda epididymis of bull after preservation at 4ºC. Totally, 23 testicles were castrated from 23 bulls (mean±standard error, age of days = 426.0±7.3, body weight (kg) = 379.7±8.4, scrotal circumference (cm) = 31.0±0.4) at Hanwoo Research Institute, NIAS, and transported to laboratory and preserved on 1, 4 and 6 days at 4 ºC. As control, epididymal spermatozoa recovery from 7 testicles was conducted after transportation to laboratory immediately. In experiment 1, we compared total number of spermatozoa among groups. Total number of spermatozoa from epididymis was not significantly on different preservation day of 0, 1, 4 and 6 which is 1778.0±304.7, 1824.8±343.9, 1228.4±91.7, 1201.8±178.6×106 cells/ml, respectively). In experiment 2, we examined spermatozoa motility and motility parameters (VCL (μm/s), VSL (μm/s), VAP (μm/s), LIN (%)) by computer assisted sperm analysis (SCA, MicroOptic) system. Percentage of motile on 0 and 1 day (88.9±5.2 and 85.8±6.1) was significantly higher than that on 4 and 6 days (32.6±6.5 and 34.3±8.25). Percentage of VCL (μm/s) on 0 and 1 day (93.5±7.6 and 83.0±14.9) was significantly higher than that on 4 and 6 days (36.6±5.1 and 39.5±5.5) (p<0.05). Percentage of VSL (μm/s) on 0 day (28.0±2.1) was significantly higher than that on 1, 4 and 6 days (20.2±3.0, 9.0±2.0 and 8.5±1.6, p<0.05). Percentage of VAP (μm/s) on 0 and 1 days (49.4±3.8 and 41.3±6.6) was significantly higher than that on 4 and 6 days (18.2±3.0 and 19.3±2.8, p<0.05). Percentage of LIN (%) on 0 day (30.7±2.6) was significantly higher than that on 4 and 6 days (23.4±2.7 and 21.1±1.0, p<0.05). Motility of spermatozoa was divided into 4 groups (fast progresive, slow progressive, non-progressive and immotile) based on WHO guideline. Percentage of fast progressive on day at 0 was significantly higher than that on 1, 4 and 6 days (0, 1, 4 and 6 days vs. 19.8±1.9, 10.2±1.1, 2.6±1.0 and 2.3±1.2%, respectively). In conclusion, cauda epididymal spermatozoa should be recovered within one day after preservation at 4 ºC to recover high quality of epididymal spermatozoa in Hanwoo bull

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.