• Journal of the Korean Society for Nondestructive Testing
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• v.5 no.1
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• pp.34-47
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• 1985

The developing country, KOREA where does not possess the natural resources for traditional energy such as oil and gas, so. The nuclear energy is the most single reliable source available for closing the energy gap. For these reason, It is inavoidable to construct the nuclear power plant and to develop technology related nuclear energy. The rate of operation in large nuclear power facilities depends upon the performance of work system through design and construction, and also the applied technology. Especially, it is the most important element that safety and reliability in operation of nuclear power plant. In view of this aspects, Nuclear power plant is performed severe examinations during preservice and inservice inspection. This study provide an automation of analysis for volumetric examination which is required to nuclear power plant components. It is composed as follows: I. Introduction II. Inservice Inspection of Nuclear Power Plant ${\ast}$ General Requirement. ${\ast}$ Principle and Methods of Ultrasonic Test. ${\ast}$ Study of Flaw Evaluation and Design of Classifying Formula for Flaws. III. Design of Automation for Flaw Evaluation. IV. An Example V. Conclusion In this theory, It is classifying the flaws, the formula of classifying flaws and the design of automation that is the main important point. As motioned the above, Owing to such as automatic design, more time could be allocated to practical test than that of evaluation of defects, Protecting against subjective bias tester by himself and miscalculation by dint of various process of computation. For the more, adopting this method would be used to more retaining for many test data and comparative evaluating during successive inspection intervals. Inspite of limitation for testing method and required application to test components, it provide useful application to flow evaluation for volumetric examination. Owing to the characteristics of nuclear power plant that is highly skill intensive industry and has huze system, the more notice should be concentrated as follows. Establishing rational operation plan, developing various technology, and making the newly designed system for undeveloped sector.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1514-1524
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• 2019

Hydrogen-steam gas mixture may be injected into containment with flow regime varying both spatially and transiently due to wall effect and pressure difference between primary loop and containment in severe accidents induced by loss of coolant accident. Preliminary CFD analysis is conducted to gain information about the helium flow regime transition process from jet to buoyancy plume for forthcoming experimental study. Physical models of impinging jet and wall condensation are validated using separated effect experimental data, firstly. Then helium transportation is analyzed with the effect of jet momentum, buoyancy and wall cooling discussed. Result shows that helium distribution is totally dominated by impinging jet in the beginning, high concentration appears near gas source and wall where jet momentum is strong. With the jet weakening, stable light gas layer without recirculating eddy is established by buoyancy. Transient reversed helium distribution appears due to natural convection resulted from wall cooling, which delays the stratification. It is necessary to concern about hydrogen accumulation in lower space under the containment external cooling strategy. From the perspective of experiment design, measurement point should be set at the height of connecting pipe and near the wall for stratification stability criterion and impinging jet modelling validation.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.107-108
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• 2005

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.185-191
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• 2020

Plant-specific analyses of an advanced reactor have been performed to assure the structural integrity of the reactor pressure vessel during transient conditions, which are expected to initiate pressurized thermal shock (PTS) events. The vessel failure probabilities from the probabilistic fracture mechanics analyses are combined with the transient frequencies to generate the through-wall cracking frequencies, which are compared to the acceptance criterion. Several sensitivity analyses are performed, focusing on the orientations and sizes of cracks, the copper content, and a flaw distribution model. The results show that the integrity of the reactor vessel is expected to be maintained for long-term operation beyond the design lifetime from the PTS perspective using the design data of the advanced reactor. Moreover, a fluence level exceeding 9×10¹⁹ n/㎠ is found to be acceptable, generating a sufficient margin beyond the design lifetime.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2909-2917
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• 2021

Following the launch of Rare Isotope Science Project in December 2011, a heavy ion accelerator complex in South Korea, named RAON, has since been designed. It includes a muon facility for muon spin rotation, relaxation, and resonance. The facility will be provided with 600 MeV and 100 kW (one-fourth of the maximum power) proton beam. In this study, the graphite target in RAON was designed to have a rotating disk shape and was cooled by radiative heat transfer. This cool-down process has the following advantages: a low-temperature gradient in the target and the absence of a liquid coolant cooling system. Monte Carlo simulations and ANSYS calculations were performed to optimize the target system in a thermally stable condition when the 100 kW proton beam collided with the target. A comparison between the simulation and experimental data was also included in the design process to obtain reliable results. The final design of the target system will be completed within 2020, and its manufacturing is in progress. The manufactured target system will be installed at the RAON in the Sindong area near Daejeon-city in 2021 to carry out verification experiments.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1416-1428
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• 2021

Pebble flow characteristics can be significantly affected by the configuration of pebble bed, especially for HTGR pebble beds. How to achieve a desired uniform flow pattern without stagnation is the top priority for reactor design. Pebbles flows inside some specially designed pebble bed with arc-shaped contraction configurations at the bottom, including both concave-inward and convex-outward shapes are explored based on discrete element method. Flow characteristics including pebble retention, residence-time frequency density, flow uniformity as well as axial velocity are investigated. The results show that the traditionally designed pebble bed with cone-shape bottom is not the most preferred structure with respect to flow pattern for reactor design. By improving the contraction configuration, the flow performance can be significantly enhanced. The flow in the convex-shape configuration featured by uniformity, consistency and less stagnation, is much more desirable for pebble bed design. In contrast, when the shape is from convex-forward to concave-inward, the flow shows more nonuniformity and stagnation in the corner although the average cross-section axial velocity is the largest due to the dominant middle pebbles.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2018.05a
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• pp.37-38
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• 2018

BIM(Building Information Modeling) in the architecture, VDC(Virtual Design and Construction) defined CIFE(Center for Integrated Facility Engineering) of Stanford university in USA, and Data-driven design definition issued by TECDOC-1284 of IAEA are doing data-level design generated by 3D CAD technology, integrating and managing related information based on the 3D model, and Using 3D models effectively during nuclear power plant life cycle. 3D model of domestic nuclear power industry is using interference review between design fields, 4D system linked 3D construction model and schedule activity, but the 3D model generated in the design phase is effectively not utilized during the construction, operation, decommissioning. therefore, This study is aimed to suggest 3D model LOD(Level of Detail) advancement method through the analysis of existing literature, 2D drawings, and 3D models throughout nuclear power plant lifecycle.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.4166-4178
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• 2021

The Control Rod Drive Mechanism (CRDM) is an essential part of the reactor, which realizes the start-stop and power adjustment of the reactor by lifting and lowering the control rod assembly. As a moving part in CRDM, the latch directly contacts with the control rod assembly, and the life of latch is closely related to the service life of the reactor. In this paper, the relationship between the life of the latch and the step stress, friction stress, and impact stress in the process of movement is analyzed, and the optimization methodology and process of latch life based on the approximate model are proposed. The design variables that affect the life of the latch are studied through the experimental design, and the optimization objective of design variables based on the latch life is established. Based on this, an approximate model of the life of the latch is built, and the multi-objective optimization of the life of the latch is optimized through the NSGA-II algorithm.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.453-462
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• 2019

Mental workload (MWL) is a major consideration for the user interface design in nuclear power plants (NPPs). However, each MWL evaluation method has its advantages and limitations, thus the evaluation and control methods based on multi-index methods are needed. In this study, fuzzy comprehensive evaluation (FCE) theory was adopted for assessment of interface designs in NPP based on operators' MWL. An evaluation index system and membership functions were established, and the weights were given using the combination of the variation coefficient and the entropy method. The results showed that multi-index methods such as performance measures (speed of task and error rate), subjective rating (NASA-TLX) and physiological measure (eye response) can be successfully integrated in FCE for user interface design assessment. The FCE method has a correlation coefficient compared with most of the original evaluation indices. Thus, this method might be applied for developing the tool to quickly and accurately assess the different display interfaces when considering the aspect of the operators' MWL.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.291-298
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• 2014

This paper concerns estimating uncertainties of the core neutronics design parameters of power reactors by direct sampling method (DSM) calculations based on the two-step McCARD/MASTER design system in which McCARD is used to generate the fuel assembly (FA) homogenized few group constants (FGCs) while MASTER is used to conduct the core neutronics design computation. It presents an extended application of the uncertainty propagation analysis method originally designed for uncertainty quantification of the FA FGCs as a way to produce the covariances between the FGCs of any pair of FAs comprising the core, or the covariance matrix of the FA FGCs required for random sampling of the FA FGCs input sets into direct sampling core calculations by MASTER. For illustrative purposes, the uncertainties of core design parameters such as the effective multiplication factor ($k_{eff}$), normalized FA power densities, power peaking factors, etc. for the beginning of life (BOL) core of Yonggwang nuclear unit 4 (YGN4) at the hot zero power and all rods out are estimated by the McCARD/MASTER-based DSM computations. The results are compared with those from the uncertainty propagation analysis method based on the McCARD-predicted sensitivity coefficients of nuclear design parameters and the cross section covariance data.