• Nuclear Engineering and Technology
• /
• v.36 no.6
• /
• pp.528-539
• /
• 2004

It is essential in commercial reactors that the safety limits imposed on the fuel pellets and fuel clad barriers, such as the linear power density (LPD) and the departure from nucleate boiling ratio (DNBR), are not violated during reactor operations. In order to accurately monitor the safety limits of current reactor states, a detailed three-dimensional (3D) core power distribution should be estimated from the in-core detector signals. In this paper, we propose a calculation methodology for detailed 3D core power distribution, using in-core detector signals and core monitoring constants such as the 3D Coupling Coefficients (3DCC), node power fraction, and pin-to-node factors. Also, the calculation method for several core safety parameters is introduced. The core monitoring constants for the real core state are promptly provided by the core design code and on-line MASTER (Multi-purpose Analyzer for Static and Transient Effects of Reactors), coupled with the core monitoring program. through the plant computer, core state variables, which include reactor thermal power, control rod bank position, boron concentration, inlet moderator temperature, and flow rate, are supplied as input data for MASTER. MASTER performs the core calculation based on the neutron balance equation and generates several core monitoring constants corresponding to the real core state in addition to the expected core power distribution. The accuracy of the developed method is verified through a comparison with the current CECOR method. Because in all the verification calculation cases the proposed method shows a more conservative value than the best estimated value and a less conservative one than the current CECOR and COLSS methods, it is also confirmed that this method secures a greater operating margin through the simulation of the YGN-3 Cycle-1 core from the viewpoint of the power peaking factor for the LPD and the pseudo hot pin axial power distribution for the DNBR calculation.

• International Journal of High-Rise Buildings
• /
• v.6 no.3
• /
• pp.217-226
• /
• 2017

This paper summarizes the structural concept and design of the "Nakanoshima Festival Tower West" in Osaka, Japan, which is 200m high and has a super-high damping system. Its superstructure is mainly composed of a central core and outer tube frames. It has a bottom truss structure at the boundary between the low-rise and mid-rise sections of the building, where the column arrangement is changed. Besides, the high-rise section of the building has a neck truss structure. These truss structures smoothly transfer the axial forces of the columns and reduce the flexural deformations induced by horizontal loads. Oil dampers with extremely high damping capacity are installed in the rigid walls named the "Big Wall Frames" of the low-rise section. Moreover, many braces and damping devices are well arranged in the center core of each story. The damping effects of these devices ensure that all structural members are remain within the elastic range and that story drifts are within 1/150 in large earthquakes. This super-high damping structure in the low-rise section is named the "Damping Layer". The whole structural system is named the "Super Damping Structure". The whole structural systems enhance the building's safety, comfort and Business Continuity Planning (BCP) under large earthquakes.

• Journal of Korean Clinical Health Science
• /
• v.5 no.2
• /
• pp.850-863
• /
• 2017

Purpose. The purpose of this study was to investigate effects of preclinical OSCE(Objective Structured Clinical Evaluation) on knowledge, confidence in their Core fundamental nursing skills and self-efficacy in nursing students. Method. The research design was a one group pretest-posttest design and it was done to assess changes in knowledge, confidence in core fundamental nursing skills and self-efficacy from pre to the post-test which was given after the OSCE. Data were collected from March 5 to April 7, 2016 from 37 nursing students who were taking the 15-hours using OSCE learning module at one Gyeongbuk-do, P-city. This practicum was composed of 4 core fundamental nursing skills and 5 other fundamental nursing skills. The knowledge consisted of a 10-item by researchers and the confidence of core fundamental nursing skills consisted of an 9-item NRS and the self-efficacy consisted of a 17-item 5-point scale and measured in both the pretest and posttest. The collected data were analyzed with SPSS IBM 20.0 program for the frequency, percentage, x2-test, and paired t-test. Rusult. The results showed that although scores of knowledge of OSCE learning module were significanlty increased from 5.22 to 7.03(t=5.30, p<.001). There were significantly increased in scores of confidence in core fundamental nursing skills from 5.13 to 7.27(t=10.01, p<.001), In the sub-scales of each core fundamental nursing skills was scored the highest. otherwise, there was no significant difference in self-efficacy(t=1.42, p=.161). Conclusions. Based on the results, this study suggests that OSCE module development activities for nursing students in nursing education-learning in order to improve nursing skills.

• The KSFM Journal of Fluid Machinery
• /
• v.17 no.6
• /
• pp.104-108
• /
• 2014

The recent Fukushima nuclear power plant accidents shows that the core make up at high RCS pressure condition is very important to prevent core melting. The core make up flow at high pressure condition should be driven by gravity force or passive forces because the AC-powered safety features are not available during a Station Black Out (SBO) accident. The reactor Coolant System (RCS) mass inventory is continuously decreased by releasing steam through the pressurizer safety valves after reactor trip during a SBO accident. The core will be melted down within 2~3 hours without core make up action by active or passive mode. In the new design concept of a Hybrid Safety Injection Tank (Hybrid SIT) both for low and high RCS pressure conditions, the low pressure nitrogen gas serves as a charging pressure for a LBLOCA injection mode, while the PZR high pressure steam provides an equalizing pressure for a high pressure injection mode such as a SBO accident. After the pressure equalizing process by battery driven initiation valve at a high pressure SBO condition, the Hybrid SIT injection water will be passively injected into the reactor downcomer by gravity head. The SBO simulation by MARS code show that the core makeup injection flow through the Hybrid SIT continued up to the SIT empty condition, and the core heatup is delayed as much.

• Journal of the Korea Society of Computer and Information
• /
• v.15 no.11
• /
• pp.1-9
• /
• 2010

As the process technology scales down, an interconnection has became a major performance constraint for multi-core processors. Recently, in order to mitigate the performance bottleneck of the interconnection for multi-core processors, a 3D integration technique has drawn quite attention. The 3D integrated multi-core processor has advantage for reducing global wire length, resulting in a performance improvement. However, it causes serious thermal problems due to increased power density. For this reason, to design efficient 3D multi-core processors, thermal-aware design techniques should be considered. In this paper, we analyze the temperature on the 3D multi-core processors in function unit level through various experiments. We also present temperature characteristics by varying application features, cooling characteristics, and frequency levels on 3D multi-core processors. According to our experimental results, following two rules should be obeyed for thermal-aware 3D processor design. First, to optimize the thermal profile of cores, the core with higher cooling efficiency should be clocked at a higher frequency. Second, to lower the temperature of cores, a workload with higher thermal impact should be assigned to the core with higher cooling efficiency.

• Journal of the Korea Academia-Industrial cooperation Society
• /
• v.17 no.10
• /
• pp.21-26
• /
• 2016

This study evaluates the magnitudes and distributions of contact tangential forces with the swaging depth of punch acting at the contact surfaces between a rotor core slot and a Cu bar during a sequential rotor core swaging process. The effects of the core slot shape on the magnitudes and distributions of the total contact forces were investigated to improve the productivity of the rotor core swaging process. Parametric elastic-plastic numerical analyses were performed using simplified two-dimensional cyclic symmetric plane strain models to evaluate the contact force distributions at the contact surfaces. The numerical analysis results show that the total contact tangential forces increased by about 55% with the adjacent Cu bar swaging process. The length of the core slot is a dominant factor in the core slot design as result of the increased total contact tangential forces during the swaging process of the rotor core.

• Journal of the Korean Society of Safety
• /
• v.33 no.5
• /
• pp.157-163
• /
• 2018

This paper describes design improvement to a research rector for safety enhancement using Probabilistic Safety Assessment (PSA). This PSA under reactor design was undertaken to assess the level of safety for the design of a research reactor and to evaluate whether it is probabilistically safe to operate and reliable to use. The scope of the PSA reported here is a Level 1 PSA, which addresses the risks associated with the core damage. The technical objectives of this study were to identify accident sequences leading to core damage and to derive design improvement from the dominant accident sequences through the sensitivity analysis. The AIMS-PSA and FTREX were used for the this PSA of the research reactor. The criterion for inclusion was all sequences with a point estimate frequency greater than a truncation value of 1.0E-14/yr. The final result indicates a point estimate of 6.79E-05/yr for the overall Core Damage Frequency (CDF) attributable to internal initiating events for the research reactor under design. Based on the dominant accident sequences from the PSA, the seven kinds of sensitivity analysis were performed and some design improvement items were derived. When the five methods to improve the safety were all applied to the reactor design and emergency operating procedure, its risk was reduced to about 1.21E-06/yr from 6.79E-05/yr. The contribution of LOCA and LOEP with high CDF were significantly reduced by the sensitivity analysis. The safety of the research reactor was well improved and the risk was reduced than before adapting the design improvement gotten from the sensitivity analysis. The present study indicated that the research reactor has the well-balanced safety in regard to each initiating event contribution to CDF. The PSA methodology is very effective to improve reactor safety in a conceptual design phase and especially, Risk-informed design(RID) is very nice way to find the deficiencies of research reactor under design and to improve the reactor safety by solving them.

• Korean Journal of Computational Design and Engineering
• /
• v.11 no.2
• /
• pp.77-87
• /
• 2006

It is necessary to change the existing design process to cope with a short life-cycle product and various customer's demands. Also a frequent design change may delay the whole design process and it will increase the unit cost of the production. New alternatives or techniques have emerged to solve the existing design problems, such as a knowledge based engineering, an intelligent CAD, a function based design, and so on. In this paper, we propose a hybrid design system with a knowledge based design methodology and a function based design technique. The knowledge based design is good at a frequent design change and the function based design is effective to extract a core design behavior. In an early design process, the system utilizes a core design behavior through the function based design process. On the other hand, the system manages complicated design issues with the knowledge based design technique in the detailed design process. We conclude that the hybrid design system can bring fair effects on implementing an efficient design environment in aspect of time and expense.

• Journal of Power Electronics
• /
• v.9 no.2
• /
• pp.180-187
• /
• 2009

When converters operate in megahertz range, the passive components and magnetic devices generate high losses. However, the eddy current issues and choices of magnetic cores significantly affect on the design stage. Apart from that, the components' reduction, miniaturization technique and frequency scaling are required as well as improvement in thermal capability, integration technique, circuit topologies and PCB layout optimization. In transformer design, the winding and core losses give great attention to the design stage. From simulation work, it is found that E-25066 material manufactured by AVX could be the most suitable core for high frequency transformer design. By employing planar geometry topology, the material can generate significant power loss savings of more than 67% compared to other materials studied in this work. Furthermore, young researchers can use this information to develop new approaches based on concepts, issues and methodology in the design of magnetic components for high frequency applications.

• 유체기계공업학회:학술대회논문집
• /
• 2005.12a
• /
• pp.503-508
• /
• 2005

An advanced Research Reactor (ARR) consists of an open-tank-type reactor assembly within a light water pool and generates thermal power of 20 MW. The thermal power is including a fission heat in the core, a fuel generated heat temporary stored in the pool, a circulating pumps generated heat and a neutron reflecting heat in the reflector vessel of the reactor. In order to remove the heat load, the primary cooling system will be installed. In this study, the conceptual design of the primary cooling system has been carried out using a design methodology of HANARO within a permissible range of safety. As results, it has been established that the conceptual design of the primary cooling system including design requirements, performance requirements, design restrictions, system descriptions and system operation to maintain the system functions.