• Nuclear Engineering and Technology
• /
• v.53 no.8
• /
• pp.2682-2695
• /
• 2021

This paper aims to evaluate the structural dynamic responses and damage/failure of the nuclear fuel reprocessing plant under the free drop impact of spent fuel cask (SFC) and fuel assembly (FA) during the on-site transportation. At the present Part I of this paper, the large-scale SFC model free drop test and the corresponding numerical simulations are performed. Firstly, a composite target which is composed of the protective structure, i.e., a thin RC plate (representing the inverted U-shaped slab in the loading shaft) and/or an autoclaved aerated concrete (AAC) blocks sacrificial layer, as well as a thick RC plate (representing the bottom slab in the loading shaft) is designed and fabricated. Then, based on the large dropping tower, the free drop test of large-scale SFC model with the mass of 3 t is carried out from the height of 7 m-11 m. It indicates that the bottom slab in the loading shaft could not resist the free drop impact of SFC. The composite protective structure can effectively reduce the damage and vibrations of the bottom slab, and the inverted U-shaped slab could relieve the damage of the AAC blocks layer dramatically. Furthermore, based on the finite element (FE) program LS-DYNA, the corresponding refined numerical simulations are performed. By comparing the experimental and numerical damage and vibration accelerations of the composite structures, the present adopted numerical algorithms, constitutive models and parameters are validated, which will be applied in the further assessment of drop impact effects of full-scale SFC and FA on prototype nuclear fuel reprocessing plant in the next Part II of this paper.

• Proceedings of the Korea Concrete Institute Conference
• /
• 2004.05a
• /
• pp.412-415
• /
• 2004

This study was performed to verify the behaviors of fiber Bragg grating (FBG) sensors attached to the containment structure in the nuclear power plant as a part of structural integrity test which demonstrates that the structural response of the non-prototype primary containment structure is within predicted limits plus tolerances when pressurized to $115\%$ of containment design pressure, and that the containment does not sustain any structural damage.

• Proceedings of the Korean Institute of Building Construction Conference
• /
• 2014.05a
• /
• pp.52-53
• /
• 2014

The Work Breakdown Structure (WBS) is a primary tool which provides a framework that defines clear scope of all deliverables throughout the project life cycle. Once the WBS is established in projects, it should allow project team members to measure and manage work performances by the WBS; further, it should provide a reference point when any work scope needs to be redefined. Based on the project information in the Progress and Performance Measurement System (PPMS) of UAE's Barakha Nuclear Power Plant (BNPP) projects, an attempt was made to develop a new WBS which provides hierarchical and systematical decomposition of the total work scope of NPP construction projects while avoiding from the preexistence concept in Korean NPP projects that the WBS is a combination of Physical Breakdown Structure (PBS) and Functional Breakdown Structure (FBS). The unique features of the new WBS are as follows: (1) defined the definition of each level of the WBS, (2) subdivided the WBS into 5 hierarchical levels, and (3) adopted globally used general coding structure. The new WBS provides a basic hierarchical structure for the project scope and can be used as a basic tool for schedule control, performance measurement, project status monitoring, and communication among project participants. In addition, by putting the Work Package (WP) under the WBS, the Earned Value Management System (EVMS) per WP can be utilized for the project.

• Journal of the Earthquake Engineering Society of Korea
• /
• v.17 no.4
• /
• pp.187-195
• /
• 2013

In this paper, firstly, acceleration-time histories were generated by varying strong motion duration in the frequency domain for application to a seismically isolated nuclear power structure, so as to examine the effects of strong motion duration on the behavior of the structure. Secondly, real recorded earthquakes were modified to match the target response spectrum based on the revised SRP 3.7.1(2007) and the modified time histories were applied to the analysis of a seismically isolated nuclear power structure. The obtained values of acceleration and displacement responses of the structure were, finally, compared with the values obtained in case of applying acceleration-time histories generated in the frequency domain to the structure.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1999.10a
• /
• pp.370.1-370
• /
• 1999

• Journal of the Earthquake Engineering Society of Korea
• /
• v.22 no.6
• /
• pp.333-343
• /
• 2018

In this paper, we study the existing results of the structure-soil-structure interaction (SSSI) effect on seismic responses of structures and summarize important parameters. The parameters considered in this study are a combination of buildings in the power block of a nuclear power plant, the characteristics of earthquake ground motions and its direction, and the characteristics embedded under the ground. Based on these parameters, the seismic analysis model of the structures in the power block of the nuclear power plant is developed and the structure-soil-structure interaction analyses are performed to analyze the influence of the parameters on the seismic response. For all analyses, the soil-structure interaction (SSI) analysis program CNU-KIESSI, which was developed to enable large-sized seismic analysis, is used. In addition, the SSI analyses is performed on individual structures and the results are compared with the SSSI analysis results. Finally, the influence of the parameters on the seismic response of the structure due to the SSSI effect is reviewed through comparison of the analysis results.

• International conference on construction engineering and project management
• /
• 2017.10a
• /
• pp.133-137
• /
• 2017

Nuclear power plants in Korea are usually built based on a duplicated model; so the project cost data of the preceding unit can be used as reference when estimating the project cost for the succeeding unit. However, since the contracting method is oriented towards the price, empirical factors such as making top-down estimations using the reverse calculation method based on the completion cost of the preceding unit is dominant. In order to develop a project cost database to resolve such problems, the detailed cost boundary of the project cost data must be categorized by project and by system. This study proposes a method to connect the code of account with the base quantities and the IAEA account, and proposes a database structure for the development of a project cost estimation system. The estimation system developed in the future is expected to utilize the proposed project cost data structure.

• Proceedings of the Korean Institute of Building Construction Conference
• /
• 2012.11a
• /
• pp.201-202
• /
• 2012

Recently, High-Strength steel reinforcement has been studied throughout the internal and external. One of the advantages using High-Strength steel reinforcement in construction is the economic effect due to the decreasing of its quantity. Also, another good effect is the increases of workability by reason of reducing the congestion. But, realistically it is not used in nuclear power plant construction site because of the restriction of design standard. The purpose of this report secures the reliability and changes the code through the performance evaluation test of the wall using the high-strength steel reinforcement in nuclear power plant.

• Journal of the Korean Society of Safety
• /
• v.36 no.5
• /
• pp.79-85
• /
• 2021

A standard engineering model that reflects the current organization system and engineering operation process of domestic nuclear power plants was developed based on the Standard Nuclear Performance Model developed by the American Nuclear Energy Association. The level 0 screen, which is the main screen of the engineering model computer system, consisted of an object tree structure, which provided information that is phased down from a higher structure level to a lower structure level (i.e., level 3). The level 1 screen provided information related to the sub-process of the engineering operation, whereas the Level 2 screen provided information related to each engineering operation activity. In addition, the Level 2 screen provided additional functions, such as linking electronic procedures/guidelines, providing electronic performance forms, and connecting legacy computer systems (such as total equipment reliability monitoring system, configuration management systems, technical information systems, risk monitoring systems, regulatory information, and electronic drawing system). This screen level increased the convenience of user's engineering tasks by implementing them. The computerization of an engineering model that connects the entire engineering tasks of an establishment enables the easy understanding of information related to the engineering process before and after the operation, and builds a foundation for the enhancement of the work efficiency and employee capacity. In addition, KHNP developed an online training module, which operates as an e-learning process, on the overview and utilization of a standard engineering model to expand the understanding of standard engineering models by plant employees and to secure competitiveness.

• Proceedings of the Earthquake Engineering Society of Korea Conference
• /
• 1997.10a
• /
• pp.90-97
• /
• 1997

The purpose of this study is to evaluate the seismic response characteristics of Wolsong nuclear power plant (NPP) structures for the Kyeongju earthquake(ML=4.3) occurred on June 26, 1997. The seismograms are obtained from five accelerographs of nuclear power plant at Wolsong, Kyeongbuk. The distance from the epicenter is about 25km. The peak acceleration (PA) due to the earthquake is 0.0235g, which is far lower value than that of design basis earthquake(DBE). The PA at the containment wall is about twice as large as that at free field. Also, the higher the accelerograph is located in, the larger the PA is measured to be From the response spectrum analysis, the dominant frequency of the response is close to 4 Hz, which is similar to the free field is poor because of contamination by high frequency waves as a result of reflection and diffraction between ground and NPP structure. We are of opinion that the accelerograph at the free field should be moved approximately twice the building dimension away from the containment structure.