• Nuclear Engineering and Technology
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• 제50권1호
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• pp.68-79
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• 2018

The Ki-Jang research reactor (KJRR), a new research reactor in Korea, is being planned to fulfill multiple purposes. In this study, as an assessment of the environmental radiological impact, we characterized the atmospheric dispersion and deposition of radioactive materials released by an unexpected incident at KJRR using the weather research and forecasting-mesoscale model interface program-California Puff (WRF-MMIF-CALPUFF) model system. Based on the reproduced three-dimensional gridded meteorological data obtained during a 1-year period using WRF, the overall meteorological data predicted by WRF were in agreement with the observed data, while the predicted wind speed data were slightly overestimated at all stations. Based on the CALPUFF simulation of atmospheric dispersion (${\chi}/Q$) and deposition (D/Q) factors, relatively heavier contamination in the vicinity of KJRR was observed, and the prevailing land breeze wind in the study area resulted in relatively higher concentration and deposition in the off-shore area sectors. We also compared the dispersion characteristics between the PAVAN (atmospheric dispersion of radioactive release from nuclear power plants) and CALPUFF models. Finally, the meteorological conditions and possibility of high doses of radiation for relatively higher hourly ${\chi}/Q$ cases were examined at specific discrete receptors.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 춘계학술발표회요약집
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• pp.253-253
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• 1999

• Journal of Mechanical Science and Technology
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• 제18권12호
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• pp.2080-2094
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• 2004

A model predictive controller is designed to control thermal power in a nuclear reactor. The basic concept of the model predictive control is to solve an optimization problem for finite future time steps at current time, to implement only the first optimal control input among the solved control inputs, and to repeat the procedure at each subsequent instant. A controller design model used for designing the model predictive controller is estimated every time step by applying a recursive parameter estimation algorithm. A 3-dimensional nuclear reactor analysis code, MASTER that was developed by Korea Atomic Energy Research Institute (KAERI), was used to verify the proposed controller for a nuclear reactor. It was known that the nuclear power controlled by the proposed controller well tracks the desired power level and the desired axial power distribution.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2017년도 춘계학술논문요약집
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• pp.33-34
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• 2017

• Nuclear Engineering and Technology
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• 제38권4호
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• pp.359-374
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• 2006

Since 1991, Korea, Canada and United States have performed the direct use of spent pressurized water reactor (PWR) fuel in the Canada deuterium uranium (CANDU) reactors (DUPIC) fuel development project. Unlike the Tandem fuel cycle, which requires a wet reprocessing, the DUPIC fuel technology can directly refabricate CANDU fuels from the PWR spent fuel and, therefore, is recognized as a highly proliferation-resistant fuel cycle technology, which can be adopted even in non-proliferation treaty countries. The Korea Atomic Energy Research Institute (KAERI) has fabricated DUPIC fuel elements in a laboratory-scale remote fuel fabrication facility. KAERI has demonstrated the fuel performance in the research reactor, and has confirmed the operational feasibility and safety of a CANDU reactor loaded with the DUPIC fuel using conventional design and analysis tools, which will be the foundation of the future practical and commercial uses of DUPIC fuel.

• 방사성폐기물학회지
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• 제16권2호
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• pp.183-193
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• 2018

현재, 한국원자력연구원은 부산 기장에 연구용 원자로(Ki-Jang Research Reactor, KJRR)를 건설 계획하고 있다. 원자로를 운영하면 중 저준위 방사성폐기물이 발생하므로 방사성 폐기물을 안전하게 처리 하는 것이 중요하다. 현재, 다양한 형태의 방사성 폐기물을 처리 할 수 있는 시멘트 고화 방법을 일반적으로 사용하고 있으며, 방사성 폐기물 처분시설 인수 기준(압축 강도, 유리수, 침수 및 침출시험 등)을 만족해야 한다. 특히, 폐기물에 함유된 방사성 세슘이 유출 될 경우 범 국제적인 문제를 야기하므로, 고화체 인수 기준 중에서 침출시험이 가장 중요한 인자이다. 시멘트 고화 방법은 다른 고화 방법 보다 공정이 간단하며 비용이 적게 들지만, 침출 저항성이 낮다. 이에 본 연구는 시멘트 고화체 세슘 침출 저항성 증진을 위하여 기장 연구용 원자로(KJRR) 모사폐액과 대표적인 세슘 흡착제인 제올라이트와 황토를 혼합하여 기장로 모의폐액 시멘트를 제조하였다. 제올라이트와 황토가 시멘트 고화체와 결합되어 있는 것을 SEM-EDS를 통하여 정량적으로 확인하였다. 침출 시험은 ANS 16.1 방법에 의해 90일동안 진행하였다. 기장로 모의폐액 시멘트의 세슘(3000 ppm)을 첨가하여 90일간의 침출시험 후 침출수의 세슘 농도 분석 결과, 제올라이트와 황토가 포함된 모의폐액 시멘트는 제올라이트와 황토를 첨가하지 않은 대조군에 비해 최대 27.90%, 21.08%의 세슘 침출 저항성 정도를 나타내는 것을 확인하였다. 또한, 제올라이트와 황토가 포함된 기장로 모의폐액 시멘트는 인수 기준(압축강도, 유리수 유무, 침수 및 침출 지수)을 통과 하는 것을 확인하였다.

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.911-919
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• 2021

In this paper, the design stress intensity values for the plate-type fuel assembly for research reactor are presented. Through a tensile test, the material properties of the cladding (aluminum alloy 6061) and structural material (aluminum alloy 6061-T6), in this case the yield and ultimate tensile strengths, Young's modulus and the elongation, are measured with the temperatures. The empirical equations of the material properties with respect to the temperature are presented. The cladding undergoes several heat treatments and hardening processes during the fabrication process. Cladding strengths are reduced compared to those of the raw material during annealing. Up to a temperature of 150 ℃, the strengths of the cladding do not significantly decrease due to the dislocations generated from the cold work. However, over 150 ℃, the mechanical strengths begin to decrease, mainly due to recrystallization, dislocation recovery and precipitate growth. Taking into account the uncertainty of the 95% probability and 95% confidence level, the design stress intensities of the cladding and structural materials are established. The presented design stress intensity values become the basis of the stress design criteria for a safety analysis of plate-type fuels.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.495-500
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• 2019

When a new fuel is developed, various mechanical properties are absolutely necessary for a safety analysis of the fuel for the licensing and prediction of its mechanical behavior during operation and accident conditions. In this paper, a mechanically equivalent surrogate plate of U-Mo dispersion fuel is presented using tungsten, substitute material of U-Mo particle. A surrogate plate, composed of tungsten/aluminum dispersion meat and aluminum alloy cladding, is manufactured with the same fabrication process with that of fuel plate except that a tungsten powder is used instead of U-Mo powder. A modal test showed that the surrogate plate and fuel plate have similar dynamic characteristics, and a tensile test demonstrated the similarity of the material property up to the yield strength range. The conducted tests proved that the surrogate tungsten plate has equivalent mechanical behaviors with that of a fuel plate, which leads to the acceptable use of a surrogate fuel assembly using tungsten/aluminum dispersion meat in various mechanical tests. The surrogate fuel assembly can be utilized for various out-of-pile characteristic tests, which are necessary for the licensing achievement of a research reactor that uses U-Mo dispersion fuel as a driver.

• 한국분말야금학회:학술대회논문집
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• 한국분말야금학회 2006년도 Extended Abstracts of 2006 POWDER METALLURGY World Congress Part 1
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• pp.474.1-474.1
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• 2006

• 한국안전학회지
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• 제30권5호
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• pp.123-130
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• 2015

Wolsong Unit 2,3&4 has been performing a containment integrity test during power operation. This test could impact to the safe operation during test. If an accident occurs during pressure dropping phase, reactor trip can be delayed because of the increased pressure difference which causes a time delay to reach the trip set-point. On the contrary, if an accident occurs during pressure increasing phase, reactor trip could be accelerated because the pressure difference to the trip set-point decrease. Point Lepreau nuclear power plant, which installed GCLMS (Gross Containment Leakage Monitoring System) in 1990, has discontinued the test since 1992 due to these adverse effects. Therefore, we evaluated the risk to obviate the GCLMS test based on PWR's ILRT (Integrated Leak Rate Test) extension methodologies. The results demonstrate that risk increase rate is not high in case of performing only ILRT test at every 5 years instead of doing GCLMS test at every 1.5 years. In addition, the result shows that GCLMS test can be removed on a risk-informed perspective since risk increasement is in acceptable area of regulatory acceptance criteria.