• Nuclear Engineering and Technology
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• v.44 no.6
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• pp.689-696
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• 2012

A small integral reactor SMART (System Integrated Modular Advanced ReacTor), being developed in Korea since late 1990s and targeted to obtaining a standard design approval by the end of 2011, is introduced. The design scope and level for design certification (DC) is well described in the U.S. NRC SECY documents published the early 1990s. However, the documents are valid for a one-step licensing process called a combined operating license (COL) by the U.S. NRC, while Korea still uses a two-step licensing process. Thus, referencing the concept of the SECY documents, we have established the design scope and level for the SMART DC using the contexts of the standard review plan (SRP). Some examples of the results and issues raised during our review are briefly presented in this paper. The same methodology will be applied to other types of reactor under development in Korea, such as future VHTR reactors.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.639-644
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• 1997

The work reported in this paper identifies the thermal-hydraulic phenomena that are expected to occur during a number of key transients in SMART(System-integrated Modular Advanced ReacTor) which is under development at KAERI. The result of this effort is based on the current design concept of SMART integral reactor. Although the design is still evolving, the preliminary phenomena Identification and Ranking Table(PIRT) has been developed based on the experts' knowledge and experience. The preliminary PIRT has been developed by consensus of KAERI expert panelists and AHP(Analytical Hierarchy Process). Preliminary PIRT developed in this paper is intended to be used to identify and integrate development areas of further experimental tests needed, thermal hydraulic models and correlations and code improvements for the safety analysis of the SMART.

• Proceedings of the Korean Information Science Society Conference
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• 2004.04b
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• pp.394-396
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• 2004

원자력 산업계에서는 원전 MMIS(Man-Machine Interface System)의 디지털 기술 적용을 위해 많은 노력을 기울이고 있고, 특히 원자력 산업의 특수성인 안전성 확보에 필요한 개발기준과 규제방법 정립에 많은 연구가 수행되고 있다. 또한 디지털 MMIS의 핵심기반기술인 고 신뢰도 소프트웨어 개발 방법론이 확립되지 못하여 소프트웨어 공통모드고장 문제, 정량적인 소프트웨어 신뢰도 보장 문제 등이 현안으로 제기되고 있다. 이와 같이 원전 MMIS의 디지털화를 성공하기 위해서는 소프트웨어의 고 신뢰도 확보가 관건이며, 고 신뢰도를 확보하기 위한 소프트웨어 개발 방법론의 정립이 절실히 요구되고 있다. 본 논문에서는 원전 소프트웨어 개발에 적응되는 규제지침을 분석하고, 일체형원자로(SMART, System-integrated Modular Advanced ReacTor) MMIS 소프트웨어 개발에 적응될 소프트웨어 개발 방법론을 제시한다.

• Proceedings of the Korea Information Processing Society Conference
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• 2004.05a
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• pp.289-292
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• 2004

원자력발전소 MMIS(Man-Machine Interface System)는 발전소 공정과 관련 장비들을 감시 및 제어하고, 필요시에 보호기능을 수행함으로써 발전소를 안전하고 신뢰성 있게 운전할 수 있도록 지원하고 있다. 그러한 MMIS의 설계에 소프트웨어기반의 컴퓨터 기술이 사용된 경우, 그 설계를 구현하기 위해 사용된 소프트웨어가 설계 및 프로그래밍 오류에 취약하여, 공통유형의 소프트웨어 오류로 인해 하드웨어로써 구축된 다중성 설계를 파기시킬 수 있기 때문에 원자력 발전소의 안전 및 안정 운전과 직결되게 된다. 또한 소프트웨어는 설계공정 결함이 일반적으로 최종 결과물에서 확인될 수 있다는 점 때문에 확인 및 검증기술을 정립하고 체계적인 적용이 필수적이다. 이에 따라 본 논문에서는 현재 설계를 진행중인 SMART(System-integrated Modular Advanced ReacTor) MMIS 소프트웨어를 개발하기 위해 적용되는 확인 및 검증 규제요건을 분석하고, 소프트웨어 개발생명주기에 따른 확인 및 검증을 체계적으로 수행하기 위한 프레임웍을 제시한다.

• Proceedings of the Korea Information Processing Society Conference
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• 2004.05a
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• pp.293-296
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• 2004

원자력 산업계에서는 원전 MMIS(Man-Machine Interface System)의 디지털 기술 적용을 위해 많은 노력을 기울이고 있고, 디지털 MMIS의 핵심기반기술인 고 신뢰도 소프트웨어 개발 방법론이 확립되지 못하여 소프트웨어 공통모드고장 문제, 정량적인 소프트웨어 신뢰도 보장 문제 등이 현안으로 제기되고 있다. 이에 따라 원자력 산업의 특수성인 안전성 확보에 필요한 개발기준과 규제방법 정립에 많은 연구가 수행되고 있다. 또한 이와 같이 원전 MMIS의 디지털화를 성공하기 위해서는 소프트웨어의 고 신뢰도 확보가 관건이며, 고 신뢰도와 품질을 확보하기 위한 소프트웨어 개발 지침의 정립이 요구되고 있다. 본 논문에서는 원전 소프트웨어 개발에 적용되는 규제 요건을 분석하고, SMART(System-integrated Modular Advanced ReacTor) MMIS 소프트웨어 개발에 적용될 소프트웨어 개발 지침을 제시한다

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1388-1395
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• 2017

The Helically coiled tube Once-Through Steam Generator (H-OTSG) is a key piece of equipment for compact small reactors. The present study developed and verified a thermal-hydraulic design and performance analysis computer code for a countercurrent H-OTSG installed in a small pressurized water reactor. The H-OTSG is represented by one characteristic tube in the model. The secondary side of the H-OTSG is divided into single-phase liquid region, nucleate boiling region, postdryout region, and single-phase vapor region. Different heat transfer correlations and pressure drop correlations are reviewed and applied. To benchmark the developed physical models and the computer code, H-OTSGs developed in Marine Reactor X and System-integrated Modular Advanced ReacTor are simulated by the code, and the results are compared with the design data. The overall characteristics of heat transfer area, temperature distributions, and pressure drops calculated by the code showed general agreement with the published data. The thermal-hydraulic characteristics of a typical countercurrent H-OTSG are analyzed. It is demonstrated that the code can be utilized for design and performance analysis of an H-OTSG.

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.1015-1024
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• 2009

The presence of a non-condensable gas can considerably reduce the level of condensation heat transfer. The non-condensable gas effect is a primary concern in some passive systems used in advanced design concepts, such as the Passive Residual Heat Removal System (PRHRS) of the System-integrated Modular Advanced ReacTor (SMART) and the Passive Containment Cooling System (PCCS) of the Simplified Boiling Water Reactor (SBWR). This study examined the capability of the Multi-dimensional Analysis of Reactor Safety (MARS) code to predict condensation heat transfer in a vertical tube containing a non-condensable gas. Five experiments were simulated to evaluate the MARS code. The results of the simulations showed that the MARS code overestimated the condensation heat transfer coefficient compared to the experimental data. In particular, in small-diameter cases, the MARS predictions showed significant differences from the measured data, and the condensation heat transfer coefficient behavior along the tube did not match the experimental data. A new method for calculating condensation heat transfer coefficient was incorporated in MARS that considers the interfacial shear stress as well as flow condition determination criterion. The predictions were improved by using the new condensation model.

• KSBB Journal
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• v.10 no.2
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• pp.159-165
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• 1995

Methyl fructoside was continuously produced in suspended bed enzyme reactor using alginate-enclosed microspheres biocatalyst which was developed for enzymatic synthesis of methyl fructoside. And the continuous operating conditions were optimized with reactor simulation in order to demonstrate a feasibility of commercialization of the continuous enzymatic production process development. The yield and productivity of methyl fructoside were as high as 47.1%o and $2g/\ell$-hr, respectively. The optimum operating conditions were pH 4.8, 30%(v/v) of methanol content and $2U/m\ell$ of enzyme activity when the initial concentration of sucrose is $0.291mo1/\ell$ at the reaction temperature of $25^{\circ}C$.

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.225-229
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• 2001

In Korea, an advanced reactor system of 330MWt power called SMART (System integrated Modular Advanced ReacTor) is being developed by KAERI to supply energy for seawater desalination as well as electricity generation. A shielding design of the SMART core with 49 CEDM is established by a two-dimensional discrete ordinates radiation transport analyses. The DORT two-dimensional discrete ordinates transport code is used to evaluate the SMART shielding designs. Three axial regions represent the SMART reactor assembly, each of which is modeled in the R-Z geometry. The BUGLE-96 library is used in the analyses, which consists of 47 neutron and 20 gamma energy groups. The results indicate that the maximum neutron fluence at the bottom of reactor vessel is $5.89 {\times} 10^{17}\;n/cm^2$ and that on the radial surface of reactor vessel is $4.49 {\times} 10^[16}\;n/cm^2$. These results meet the requirement, $1.0 {\times} 10^{20}\;n/cm^2$, in 10 CFR 50.61 and the integrity of SMART reactor vessel during the lifetime of the reactor is confirmed.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.151-156
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• 2001

Experimental frictional and wear characteristics of silicon graphite materials is studied in this paper. Those specimens are lubricated with high temperature and highly pressurized water to simulate the same operating condition for the journal bearing and the thrust bearing on the main coolant pump bearing in the newly developing nuclear reactor named SMART(System-integrated Modular Advanced ReacTor). Operating condition of the bearings is realized by the tribometer and the autoclave. Friction coefficient and wear loss are analyzed to choose the best silicon graphite material. Pin on plate test specimens are used and coned disk springs are used to control the applied force on the specimens. Wear loss ana wear width are measured by a precision balance and a micrometer. The friction force is measured by the strain gauge which can be used under high temperature and high pressure. Three kinds of silicon graphite materials are examined and compared with each other, and each material shows similar but different results on frictional and wear characteristics.