• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.65-66
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• 2005

• 한국초전도학회:학술대회논문집
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• v.12
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• pp.82-82
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• 2002

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2262-2275
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• 2022

The Korea Atomic Energy Research Institute (KAERI) studied the sodium-water reaction (SWR) minimized steam generator for the safety of the sodium-cooled fast reactor (SFR), and selected the copper bonded steam generator (CBSG) as the optimal concept. This paper introduces the conceptual design of the CBSG and the development of the CBSG sizing analyzer (CBSGSA). The CBSG consists of multiple heat transfer modules with a crossflow heat transfer configuration where sodium flows horizontally and water flows vertically. The heat transfer modules are stacked along a vertical direction to achieve the targeted large heat transfer capacity. The CBSGSA code was developed for the thermal-hydraulic analysis of the CBSG in a multi-pass crossflow heat transfer configuration. Finally, we conducted a preliminary sizing and rating analysis of the CBSG for the trans-uranium (TRU) core system using the CBSGSA code proposed by KAERI.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1847-1861
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• 2020

A design study was performed to improve the limit aseismic performance (LSP) of a primary heat transport system (PHTS) pump. This pump is part of the primary equipment of a prototype generation IV sodium-cooled fast reactor (PGSFR). The LSP is the maximum allowable seismic load that still ensures structural integrity. To calculate the LSP of the PHTS pump, a structural analysis model of the pump was developed and its dynamic characteristics were obtained by modal analysis. The floor response spectrum (FRS) initiated from a safety shutdown earthquake (SSE), 0.3 g, was applied to the support points of the PHTS pump, and then the seismic induced stresses were calculated. The structural integrity was evaluated according to the ASME code, and the LSP of the PHTS pump was calculated from the evaluation results. Based on the results of the modal analysis and LSP of the PHTS pump, design parameters affecting the LSP were selected. Then, ways to improve the LSP were proposed from sensitivity analysis of the selected design variables.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.4042-4051
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• 2021

In this work, we applied a multilevel modeling technique to estimate the diametral creep in the pressure tubes of Korean Canada Deuterium Uranium (CANDU) units. Data accumulated from in-service inspections were used to develop the model. To confirm the strength of the multilevel models, a 2-level multilevel model considering the relationship between channels for a CANDU unit was compared with existing linear models. The multilevel model exhibited a very robust prediction accuracy compared to the linear models with different data pooling methods. A 3-level multilevel model, which considered individual bundles, channels, and units, was also implemented. The influence of the channel installation direction was incorporated into the three-stage multilevel model. For channels that were previously measured, the developed 3-level multilevel model exhibited a very good predictive power, and the prediction interval was very narrow. However, for channels that had never been measured before, the prediction interval widened considerably. This model can be sufficiently improved by the accumulation of more data and can be applied to other CANDU units.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.240-252
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• 2015

In the original licensing application for the prototype fast-breeder reactor, MONJU, the event progression during an unprotected loss of flow (ULOF), which is one of the technically inconceivable events postulated beyond design basis, was evaluated. Through this evaluation, it was confirmed that radiological consequences could be suitably limited even if mechanical energy was released. Following the Fukushima-Daiichi accident, a new nuclear safety regulation has become effective in Japan. The conformity of MONJU to this new regulation should hence be investigated. The objectives of the present study are to conduct a preliminary evaluation of ULOF for MONJU, reflecting the knowledge obtained after the original licensing application through CABRI experiments and EAGLE projects, and to gain the prospect of in-vessel retention for the conformity of MONJU to the new regulation. The preliminary evaluation in the present study showed that no significant mechanical energy release would take place, and that thermal failure of the reactor vessel could be avoided by the stable cooling of disrupted-core materials. This result suggests that the prospect of in-vessel retention against ULOF, which lies within the bounds of the original licensing evaluation and conforms to the new nuclear safety regulation, will be gained.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.572-594
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• 2016

As one of the Gen-IV nuclear energy systems, a sodium-cooled fast reactor (SFR) is being developed at the Korea Atomic Energy Research Institute. As a long-term national research project, advanced radiation resistant oxide dispersion strengthened steel (ARROS) is being developed as an in-core fuel cladding tube material for a SFR in the future. In this paper, the current status of ARROS development is reviewed and its future prospective is discussed.

• Journal of the Korean Ceramic Society
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• v.42 no.8 s.279
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• pp.575-581
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• 2005

SiC-based ceramics are considered as candidate materials for the advanced nuclear energy systems such as the generation IV reactors and the fusion reactors due to their excellent high-temperature strength and irradiation resistance. The advanced nuclear energy systems and their main components adopting ceramic composites were briefly reviewed. A novel fabrication method of $SiC_f/SiC$ composites by introducing SiC whiskers was also described. In addition, the charged-particle irradiation ($Si^{2+}$ and $H^{+}$ ion) into CVD SiC was carried out to simulate the severe environments of the advanced nuclear reactors. SiC whiskers grown in the fiber preform increased the matrix infiltration rate by more than $60\%$ compared to the conventional CVI process. The highly crystalline and pure SiC showed little degradation in hardness and elastic modulus up to a damage level of 10 dpa at $1000^{\circ}C$.

• Journal of the Korea Institute of Military Science and Technology
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• v.25 no.6
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• pp.580-585
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• 2022

In this study, the high-temperature dielectric constant of Si₃N₄ ceramics, a representative non-oxide-based radome material, was evaluated and the cause of the dielectric constant change was analyzed in relation to the oxidation behavior. The dielectric constant of Si₃N₄ ceramics was 7.79 at room temperature, and it linearly increased as the temperature increased, showing 8.42 at 1,000 ℃. As results of analyzing the microstructure and phase for the Si₃N₄ ceramics before and after heat-treatment, it was confirmed that oxidation did not occur at all or occurred only on the surface at a very insignificant level below 1,000 ℃. Based on this, it is concluded that the increase in the dielectric constant according to the temperature increase of Si₃N₄ ceramics is irrelevant to the oxidation behavior and is only due to the activation of charge polarization.

• Proceedings of the KIEE Conference
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• 2008.10b
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• pp.540-541
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• 2008

원자력 발전소 격납구조(containment) 내에 설치되는 센서, 구동기(actuator) 및 설비는 원전의 안전운전과 함께 방사능 누출사고와 같은 중대사고(severe accident)를 예방하기 위한 것이다. 격납구조 내부는 Category I 등급으로 분류되며, 격납구조 내부에 설치되는 센서, 구동기, 기기 및 통신망은 IEEE Std. 323-1974에서 정의하는 극한환경(harsh environment) 요건에서 생존할 수 있는 내환경성이 요구된다. 이러한 엄격한 내환경성 요건으로 인해 일반 산업의 IT 기반 센서통신망이 원전 격납건물 내부에는 적용되지 않고 있다. 최근에 이르러 독일을 중심으로 신규로 건설 중이거나 계획 중인 원전에서는 일반 산업의 IT 기반 센서 통신망 적용이 검토되고 있다. 본 논문에서는 IT 기반의 첨단 센서 통신망 기술을 격납구조내부와 같은 극한 환경에 적용하기 위한 방안을 제시하고자 한다. 정상운전중의 원전 격납 건물 내부의 환경(온도, 감마선, 습도) 특성과 중대 사고를 가정한 DBA (설계 기준사고) 요건에서의 환경 특성을 조사하였다. 또한 설계기준사고에서 정의한 감마선 조사 환경에서 통신 시스템의 생존성을 실험하였다. 이를 토대로 격납구조내부의 원전 극한 환경 통신망의 개선방안을 제시하고자 한다.