• Journal of Information Technology Applications and Management
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• v.11 no.4
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• pp.35-47
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• 2004

The nuclear design analysis requires time-consuming and erroneous model-input preparation. code run. output analysis and quality assurance process. To reduce human effort and improve design quality and productivity. Innovative Design Processor (IDP) is being developed. Two basic principles of IDP are the document-oriented desigll and the web-based design. The document-oriented design is that. if the designer writes a design document called active document and feeds it to a special program. the final document with complete analysis. table and plots is made automatically. The active documents can be written with Microsoft Word or created automatically on the web. which is another framework of IDP. Using the proper mix-up of server side and client side programming under the LAMP (Linux/Apache/MySQL/PHP) environment. it e design process on the web is modeled as a design wizard style so that even a novice designer makes the design document easily. This automation using the IDP is now being implemented for all the reload design of Korea Standard Nuclear Power Plant (KSNP) type PWRs. The introduction of this process will allow large reduction in all reload design efforts of KSNP and provide a platform for design and R&D tasks of KNFC.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.23 no.1
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• pp.49-55
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• 2013

A reactor vessel internals comprehensive vibration assessment program(RVI CVAP) of an advanced power reactor 1400(APR1400) is being performed as a non-prototype category-2 type of reactor based on the US nuclear regulatory commission regulatory guide(NRC RG) 1.20. The aim of this paper is to present the results of structural response analysis and measuring locations of a upper guide structure(UGS) assembly of the APR1400 reactor. The analysis results of the UGS assembly show that the specified integrity levels meet the design acceptance criteria. Also, the measuring locations are determined by the analysis results of the UGS assembly and selection criteria of previous study. These analysis results and measuring locations will be used as a guide to design a measurement system for the APR1400 RVI CVAP.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.1
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• pp.23-34
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• 2004

To validate the previous conceptual design of cover system, construction of the engineered barrier test facility is completed and the performance tests of the disposal cover system are conducted. The disposal test facility is composed of the multi-purpose working space, the six test cells and the disposal information space for the PR center. The dedicated detection system measures the water content, the temperature, the matric potential of each cover layer and the accumulated water volume of lateral drainage. Short-term experiments on the disposal cover layer using the artificial rainfall system are implemented. The sand drainage layer shows the satisfactory performance as intended in the design stage. The artificial rainfall does not affect the temperature of cover layers. It is investigated that high water infiltration of the artificial rainfall changes the matric potential in each cover layer. This facility is expected to increase the public information about the national radioactive waste disposal program and the effort for the safety of the planned disposal facility.

• Nuclear Engineering and Technology
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• v.41 no.4
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• pp.447-454
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• 2009

This paper is a technical review of the research and development results of the Korea Nuclear Instrumentation and Control System (KNICS) project and Nu-Tech 2012 program. In these projects man-machine interface system architecture, two digital platforms, and several control and protection systems were developed. One platform is a Programmable Logic Controller (PLC) for a digital safety system and another platform is a Distributed Control System (DCS) for a non-safety control system. With the safety-grade platform PLC, a reactor protection system, an engineered safety feature-component control system, and reactor core protection system were developed. A power control system was developed based on the DCS. A logic alarm cause tracking system was developed as a man-machine interface for APR1400. Also, Integrated Performance Validation Facility (IPVF) was developed for the evaluation of the function and performance of developed I&C systems. The safety-grade platform PLC and the digital safety system obtained approval for the topical report from the Korean regulatory body in February of 2009. A utility and vendor company will determine the suitability of the KNICS and Nu- Tech 2012 products to apply them to the planned nuclear power plants.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.469-474
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• 1997

To provide tile information ell severe accident progression is very important for advanced or new type of nuclear power plant (NPP) design. A parametric study, therefore was performed to investigate the effect of thermal hydraulic design parameters ell severe accident progression of pressurized water reactors (PWRs), Nine parameters, which are considered important in NPP design or severe accident progression, were selected among the various thermal hydraulic design parameters. The backpropagation neural network (BPN) was used to determine parameters, which might more strongly affect the severe accident progression, among mile parameters. For training. different input patterns were generated by the latin hypercube sampling (LHS) technique and then different target patterns that contain core uncovery time and vessel failure time were obtained for Young Gwang Nuclear (YGN) Units 3&4 using modular accident analysis program (MAAP) 3.0B code. Three different severe accident scenarios, such as two loss of coolant accidents (LOCAs) and station blackout(SBO), were considered in this analysis. Results indicated that design parameters related to refueling water storage tank (RWST), accumulator and steam generator (S/G) have more dominant effects on the progression of severe accidents investigated, compared to tile other six parameters.

• Journal of the Earthquake Engineering Society of Korea
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• v.17 no.2
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• pp.53-59
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• 2013

Several researches have been studied to enhance the seismic performance of nuclear power plants (NPPs) by application of seismic isolation. If a seismic base isolation system is applied to NPPs, seismic performance of nuclear power plants should be reevaluated considering the soil-structure interaction effect. The seismic fragility analysis method has been used as a quantitative seismic safety evaluation method for the NPP structures and equipment. In this study, the seismic performance of an isolated NPP is evaluated by seismic fragility curves considering the soil-structure interaction effect. The designed seismic isolation is introduced to a containment building of Shin-Kori NPP which is KSNP (Korean Standard Nuclear Power Plant), to improve its seismic performance. The seismic analysis is performed considering the soil-structure interaction effect by using the linearized model of seismic isolation with SASSI (System for Analysis of Soil-Structure Interaction) program. Finally, the seismic fragility is evaluated based on soil-isolation-structure interaction analysis results.

• Nuclear Engineering and Technology
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• v.44 no.5
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• pp.543-550
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• 2012

The decrease in the fracture toughness of ferritic steels in a reactor pressure vessel is an important factor in determining the lifetime of a nuclear power plant. A surveillance program has been in place in Korea since 1979 to assess the structural integrity of RPV steels. In this work, the surveillance data were collected and analyzed statistically in order to derive the empirical relationship between the embrittlement and strengthening of irradiated reactor pressure vessel steels. There was a linear relationship between the yield strength change and the transition temperature shift change at 41 J due to irradiation. The proportional coefficient was about $0.5^{\circ}C$/MPa in the base metals (plate/forgings). The upper shelf energy decrease ratio was non-linearly proportional to the yield strength change, and most of the data lay along the trend curve of the US results. The transition regime temperature interval, ${\Delta}T_T$, was less than the US data. The overall change from irradiation was very similar to the US results. It is expected that the results of this study will be applied to basic research on the multiscale modeling of the irradiation embrittlement of RPV materials in Korea.

• The Korean Journal of Nuclear Medicine
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• v.37 no.4
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• pp.207-218
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• 2003

Myocardial perfusion imaging has been increasingly used to provide prognostic data and guidance on the choice of appropriate management of patients with known or suspected coronary artery disease. The electrocardiogram gated myocardial SPECT program is corning into wide use with an advent of $^{99m}Tc-labeled$ tracers and an improvement of SPECT machines. The gated technique permits measurement of important cardiac prognostic indicators without any further discomforts or radiation burden in patients underwent standard myocardial perfusion SPECT. In addition, gated study significantly improves diagnostic yield by reducing the number of borderline interpretations and could find myocardial stunning and viable myocardium. Gated single photon emission computed tomography (SPECT) imaging allows the automated calculation of end-diastolic volume, end-systolic volume, ejection fraction, myocardial mass and the assessment of regional wall motion and thickening, and it have dramatically improved assessment of coronary artery disease in routine nuclear practice. This allows the simultaneous assessment of both perfusion and function within the same acquisition, and serves as a cost-effective technique for providing more diagnostic data with fewer diagnostic tests. Because the diagnostic and prognostic power derived from knowledge of left ventricular function can be added to that provided by assessing myocardial perfusion, gated SPECT imaging has rapidly gained widespread acceptance and is now used on a routine clinical basis in a growing number of laboratories, including South Korea. The gated SPECT technique for measurement of left ventricular parameters has been validated against a variety of well established techniques. In this work, overview of gated myocardial perfusion SPECT focus on functional parameters is presented.

• Journal of the Korean Society of Safety
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• v.27 no.6
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• pp.14-19
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• 2012

The Helium cooled ceramic reflector(HCCR) test blanket module(TBM) has been designed and developed to participate the ITER(International Thermonuclear Experimental Reactor) test blanket program in Korea. The TBM was one of the main objectives for developing ITER for proving the tritium self-sufficiency and the heat transfers to produce the electricity with the breeding blanket concept. Among the TBM components, the first wall(FW) was the most important component in safety since it was directly faced a high level of a heat and fast neutrons from the plasma side and could protect the others components inside TBM. In this paper, the FW has been designed through the thermo-mechanical analysis considering ITER operation conditions. With the developed simple models, the stress limit analysis based on RCC-MR code which is the nuclear power plant design codes in France was evaluated for the allowable design criteria. The results showed that the designed FW model satisfied $1.5S_m$ or $3S_m$ of the allowable stress($S_m$) in RCC-MR code at the maximum stress region in the FW.

• Proceedings of the KIEE Conference
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• 2007.07a
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• pp.614-615
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• 2007

There are roughly a hundred types of cables in power plants. The distribution of circuits in a nuclear plant is comprised of 20% instrument cables, 61% control cables, 13% AC power cables, 1% DC power cables, and 5% communication lines. In the nuclear power plant, medium voltage cables are generally included in the scope of systems reviewed for safety and are included in a plant's maintenance program. Medium voltage cables provide power to many critical components in plants, including feed water pumps, circulating water pumps, and condensate pumps. Among these cables, high temperature sections of cables feeding electrical power to the circulating water pump and the condenser extraction pump were found. The evaluation for these cables is performed to find the maximum allowable current and temperature. The result shows that the load current flowed about 85% of the allowable current ampacity, and the temperature of conductor at full load current did not exceed the limited temperature. Therefore, existing cables for circulating water pump and condenser extraction pump system are going to be used during design life.