• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.2
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• pp.113-119
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• 2011

Safe operation and maintenance of engineered dry storage systems for spent fuel from nuclear power plants basically depends on adequately adopted design requirements. The most important design target of the system are those which provide the necessary assurances that spent fuel can be received, handled, stored and retrieved without undue risk to health and safety of workers and the public. To achieve these objectives, the design of the system incorporates features to remove spent fuel residual heat, to provide for radiation protection, and to maintain containment over the lifespan of the system as specified in the design specifications. The features also provide for all possible anticipated operational occurrences and design basis events in accordance with the design basis as guided by the designated regulations. The general performance requirements of a projected storage system are introduced in this paper. The storage system is designed to store fuel assemblies in associated with designated regulatory requirements. Small increases/decreases in maximum burnup can be adjusted with cooling time. These variations are compensated for by a corresponding small site-specific increase/decrease in the design basis-cooling period, as long as the maximum heat load and radioactivity of loaded fuel assemblies are met. Generic design basis events considered for the storage system are summarized. Shielding and radiological requirements along with mechanical and structural are derived in this study.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.1
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• pp.29-36
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• 2019

High-level radioactive waste (HLW) such as spent fuel is inevitably produced when nuclear power plants are operated. A geological repository has been considered as one of the most adequate options for the disposal of HLW, and it will be constructed in host rock at a depth of 500~1,000 meters below ground level with the concept of an engineered barrier system (EBS) and a natural barrier system. The compacted bentonite buffer is one of the most important components of the EBS. As the compacted bentonite buffer is located between disposal canisters with spent fuel and the host rock, it can restrain the release of radionuclides and protect canisters from the inflow of groundwater. Because of inflow of groundwater into the compacted bentonite buffer, it is essential to investigate soil-water characteristic curves (SWCC) of the compacted bentonite buffer in order to evaluate the entire safety performance of the EBS. Therefore, this paper conducted laboratory experiments to analyze the SWCC for a Korean Ca-type compacted bentonite buffer considering dry density, confined or unconfined condition, and drying or wetting path. There was no significant difference of SWCC considering dry density under unconfined condition. Furthermore, it was found that there was higher water suction in unconfined condition that in confined condition, and higher water suction during drying path than during wetting path.

• Journal of Platform Technology
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• v.9 no.3
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• pp.52-62
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• 2021

Quality control is critical at manufacturing sites and is key to predicting the risk of quality defect before manufacturing. However, the reliability of manual quality control methods is affected by human and physical limitations because manufacturing processes vary across industries. These limitations become particularly obvious in domain areas with numerous manufacturing processes, such as the manufacture of major nuclear equipment. This study proposed a novel method for predicting the risk of quality defects by using natural language processing and machine learning. In this study, production data collected over 6 years at a factory that manufactures main equipment that is installed in nuclear power plants were used. In the preprocessing stage of text data, a mapping method was applied to the word dictionary so that domain knowledge could be appropriately reflected, and a hybrid algorithm, which combined n-gram, Term Frequency-Inverse Document Frequency, and Singular Value Decomposition, was constructed for sentence vectorization. Next, in the experiment to classify the risky processes resulting in poor quality, k-fold cross-validation was applied to categorize cases from Unigram to cumulative Trigram. Furthermore, for achieving objective experimental results, Naive Bayes and Support Vector Machine were used as classification algorithms and the maximum accuracy and F1-score of 0.7685 and 0.8641, respectively, were achieved. Thus, the proposed method is effective. The performance of the proposed method were compared and with votes of field engineers, and the results revealed that the proposed method outperformed field engineers. Thus, the method can be implemented for quality control at manufacturing sites.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.4
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• pp.319-327
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• 2010

A nuclear plant ESF ACS simulator was designed, built, and verified to perform experiment related to ESF ACS of nuclear power plants. The dimension of 3D CAD model was based on drawings of the main control room(MCR) of Yonggwang units 5 and 6. The CFD analysis was performed based on the measurement of the actual flow rate of ESF ACS. The air flowing in ACS was assumed to have $30^{\circ}C$ and uniform flow. The flow rate across the HEPA filter was estimated to be 1.83 m/s based on the MCR ACS flow rate of 12,986 CFM and HEPA filter area of 9 filters having effective area of $610{\times}610mm^2$ each. When MCR ACS was modeled, air flow blocking filter frames were considered for better simulation of the real ACS. In CFD analysis, the air flow rate in the lower part of the active carbon adsorber was simulated separately at higher than 7 m/s to reflect the measured value of 8 m/s. Through the CFD analyses of the ACSes of fuel building emergency ventilation system, emergency core cooling system equipment room ventilation cleanup system, it was confirmed that all three EFS ACSes can be simulated by controlling the flow rate of the simulator. After the CFD analysis, the simulator was built in nuclear grade and its reliability was verified through air flow distribution tests before it was used in main tests. The verification result showed that distribution of the internal flow was uniform except near the filter frames when medium filter was installed. The simulator was used in the tests to confirm the revised contents in Reg. Guide 1.52 (Rev. 3).

• Ocean and Polar Research
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• v.28 no.4
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• pp.385-393
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• 2006

Screening performance of the existing intake screens (drum and travelling screen) on mass impingement of marine animals, a euphausiid, Euphausia pacifica and a scyphozoan medusae, Aurelia aurita that have often clogged intake screens of the Uljin Nuclear Power Plant, was tested. The maximum tolerable densities of marine animals in the inflowing seawater upon the screen were estimated with two different approaches. First the maximum density of jellyfish was calculated from (1) passing amount of seawater per unit time through the screens and (2) the covered area of animals on the screens clogged. The maximum density of krill tolerable in the drum screen was cited from a simulated record of Uljin NPP, then those in the travelling screens were also calculated using the data of drum screen and ratio of seawater amount passing through the screens under the condition of 0.5m water column (W.C.) of the differential pressure (AP) produced by screens, an established permissible limit of ${\Delta}P$. Secondly, the screening performances were also tested by hydrodynamic measurements with various screen models in a circulating water channel equipped with a speed-controlling pump and a differential pressure gauge. From the first approach, the maximum tolerable densities of drum and travelling screen were calculated as 2.0 and $1.5ind/m^3$ for the Jellyfish and 900 and $680ind./m^3$ for the euphausiid, respectively. These densities estimated from the second approach were 2.1 and $0.8ind/m^3$ for the jellyfish and 1059 and $504ind/m^3$ for the euphausiid, respectively. These estimates were compared with the data from historic clogging events to evaluate the practical performance of these intake screens. The comparisons suggest a newly improved intake-screen of which performance should be at least seven times (approximately) better than the existing ones ior the krill and 3.2 times for the jellyfish, respectively, for preventing mass impingement, and for maintaining the condition of the differential pressure between the screens below 0.3 m W.C.

• Journal of the Korean Society for Nondestructive Testing
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• v.29 no.4
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• pp.374-382
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• 2009

Steam generator(SG) tube, as a barrier isolating primary to the secondary coolant system of nuclear power plants(NPP), must maintain the structural integrity far the public safety and its efficient power generation capacity. And SG tubes bearing defects must be timely detected and taken repair measures if needed. For the accomplishment of these objectives, SG tubes have been periodically examined by eddy current testing(ECT) on the basis of administrative notices and intensified SG management program(SGMP). Stress corrosion cracking(SCC) on the SG tubes is not easily detected and even missed since it has lower signal amplitude and other disturbing factors against its detection. However once SCC is developed, that can cause detrimental affects to the SG tubes due to its rapid propagation rate. Accordingly SCC is categorized as prime damage mechanism challenging the soundness of the SG tubes. In this study, reproduced EDM notch specimens are examined for the detectability and quantitative characterization of the axial ODSCC by +PT MRPC probe, containing pancake, +PT and shielded pancake coils apart in a single plane around the circumference. The results of this study are assumed to be applicable fur providing key information of engineering evaluation of SCC and improvement of confidence level of ECT on SG tubes.

• Computational Structural Engineering
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• v.7 no.3
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• pp.95-103
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• 1994

The safety related components in the nuclear power plant should be designed to withstand the seismic load. Among these components the integrity of reactor internals under earthquake load is important in stand points of safety and economics, because these are classified to Seismic Class I components. So far the modelling methods of reactor internals have been investigated by many authors. In this paper, the dynamic behaviour of reactor internals of Yong Gwang 1&2 nuclear power plants under SSE(Safe Shutdown Earthquake) load is analyzed by using of the simpled Global Beam Model. For this, as a first step, the characteristic analysis of reactor internal components are performed by using of the finite element code ANSYS. And the Global Beam Model for reactor internals which includes beam elements, nonlinear impact springs which have gaps in upper and lower positions, and hydrodynamical couplings which simulate the fluid-filled cylinders of reactor vessel and core barrel structures is established. And for the exciting external force the response spectrum which is applied to reactor support is converted to the time history input. With this excitation and the model the dynamic behaviour of reactor internals is obtained. As the results, the structural integrity of reactor internal components under seismic excitation is verified and the input for the detailed duel assembly series model could be obtained. And the simplicity and effectiveness of Global Beam Model and the economics of the explicit Runge-Kutta-Gills algorithm in impact problem of high frequency interface components are confirmed.

• Korean Journal of Food Science and Technology
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• v.36 no.1
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• pp.183-187
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• 2004

Radioactivity in foodstuffs was surveyed for reference in amending regulation on the maximum permitted levels of radioactive contamination of foodstuffs. Most domestic and imported (?) foodstuffs were sampled, some domestic items collected around nuclear power plants to compare site-specific contamination. The collected samples were dried and ashed. Radioactivity in foodstuffs was measured using HPGe gamma spectrometer, Cs-137 activity ranged from 0.025-0.053, 0.045-0.500, 0.062-0.105, 0.025-1.151, 0.021-0.145 and 0.046-0.155 Bq/kg-fresh in cereals, pulses, mot vegetables (potato), ginsengs, meat, and marine products, respectively, with imported dried ginseng showing the highest radioactivity, Results reveal radioactivity in foodstuffs collected in 2002 is far below the maximum permitted levels of 370 Bq/kg. No significant differences were observed in radioactivity among sampling sites and between domestic and imported foodstuffs.

• Economic and Environmental Geology
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• v.56 no.4
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• pp.421-433
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• 2023

The final disposal of spent nuclear fuel(SNF) from nuclear power plants takes place in a deep geological repository. The metal canister encasing the SNF is made of cast iron and copper, and is engineered to effectively isolate radioactive isotopes for a long period of time. The SNF is further shielded by a multi-barrier disposal system comprising both engineering and natural barriers. The deep disposal environment gradually changes to an anaerobic reducing environment. In this environment, sulfide is one of the most probable substances to induce corrosion of copper canister. Stress-corrosion cracking(SCC) triggered by sulfide can carry substantial implications for the integrity of the copper canister, potentially posing a significant threat to the long-term safety of the deep disposal repository. Sulfate can exist in various forms within the deep disposal environment or be introduced from the geosphere. Sulfate has the potential to be transformed into sulfide by sulfate-reducing bacteria(SRB), and this converted sulfide can contribute to the corrosion of the copper canister. Bentonite, which is considered as a potential material for buffering and backfilling, contains oxidized sulfate minerals such as gypsum(CaSO4). If there is sufficient space for microorganisms to thrive in the deep disposal environment and if electron donors such as organic carbon are adequately supplied, sulfate can be converted to sulfide through microbial activity. However, the majority of the sulfides generated in the deep disposal system or introduced from the geosphere will be intercepted by the buffer, with only a small amount reaching the metal canister. Pyrite, one of the potential sulfide minerals present in the deep disposal environment, can generate sulfates during the dissolution process, thereby contributing to the corrosion of the copper canister. However, the quantity of oxidation byproducts from pyrite is anticipated to be minimal due to its extremely low solubility. Moreover, the migration of these oxidized byproducts to the metal canister will be restricted by the low hydraulic conductivity of saturated bentonite. We have comprehensively analyzed and summarized key research cases related to the presence of sulfates, reduction processes, and the formation and behavior characteristics of sulfides and pyrite in the deep disposal environment. Our objective was to gain an understanding of the impact of sulfates and sulfides on the long-term safety of high-level radioactive waste disposal repository.

• Tunnel and Underground Space
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• v.19 no.6
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• pp.490-497
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• 2009

The horizontal and vertical response spectra using the observed ground motion from the recent 5 macro earthquakes were analysed and then were compared to both the seismic design response spectra(Reg Guide 1.60), applied to the domestic nuclear power plants, and the Korean Standard Design Response Spectrum for general structures and buildings(1997). 74 horizontal and 89 vertical observed ground motions, without considering soil types, were used for normalization with respect to the peak acceleration value of each ground motion. The results showed that the horizontal MPOSD(Mean Plus One Sigma Standard Deviation) response spectra revealed much higher values for the whole frequency bands above 1 Hz than Reg. Guide(1.60). For the vertical response spectra, the results showed slightly higher than just between 7 and 8 Hz frequency band. The results were also compared to the Korean Standard Response Spectrum for the 3 different soil types and showed that the horizontal MPOSD response spectra revealed much higher values for the whole periods below 2 second(0.5 Hz) than those of SE soil type. The vertical response spectra showed similar to the values of the Korean Standard Response Spectrum of SD soil type. These spectral values dependent on frequency could be related to characteristics of the domestic crustal attenuation and the effect of each site amplification. However, through the qualitative improvements and quantitative enhancement of the observed ground motions, the conservation of horizontal seismic design response spectrum should be considered more significantly for the whole frequency bands above the 1 Hz.