• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2003년도 가을 학술논문집
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• pp.38-42
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• 2003

가압경수형 원자력발전소 일차계통에서 발생되는 방사성 부식생성물(크러드)은 원자력발전소 작업종사자 피폭의 주요원인이다. 또한, 최근 원자력발전소의 장주기운전 추세에 따라 장기간 노심에 침적된 방사성 부식생성물은 hideout 현상으로 노심의 출력에 영향을 주는 축방향이상출력(AOA) 현상의 원인이 되고 있다. 크러드의 주요 성분은 마그네타이트, 니켈페라이트, 코발트페라이트가 주를 이루며, 이러한 산화물 형태는 강자성의 자기적 성질을 가지고 있다. 따라서, 전자석과 영구자석의 적절한 배치를 통하여 자기장을 발생시켜 크러드를 제거하는 필터 개발을 위해 개념 설계를 하였다. 기존의 필터와 달리 유체의 흐름을 방해하지 않아 압력저하 현상이 발생하지 않고, 연속적으로 사용할 수 있는 장점이 있다. 크러드 제거 기술의 하나로써 활용될 수 있을 것으로 기대된다.

• Nuclear Engineering and Technology
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• 제43권2호
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• pp.133-140
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• 2011

In this paper, a deformation measurement technology for nuclear fuel rods is proposed. The deformation measurement system includes a high-definition CMOS image sensor, a lens, a semiconductor laser line beam marker, and optical and mechanical accessories. The basic idea of the proposed deformation measurement system is to illuminate the outer surface of a fuel rod with a collimated laser line beam at an angle of 45 degrees or higher. For this method, it is assumed that a nuclear fuel rod and the optical axis of the image sensor for observing the rod are vertically composed. The relative motion of the fuel rod in the horizontal direction causes the illuminated laser line beam to move vertically along the surface of the fuel rod. The resulting change of the laser line beam position on the surface of the fuel rod is imaged as a parabolic beam in the high-definition CMOS image sensor. An ellipse model is then extracted from the parabolic beam pattern. The center coordinates of the ellipse model are taken as the feature of the deformed fuel rod. The vertical offset of the feature point of the nuclear fuel rod is derived based on the displacement of the offset in the horizontal direction. Based on the experimental results for a nuclear fuel rod sample with a formation of surface crud, an inspection resolution of 50 ${\mu}m$ is achieved using the proposed method. In terms of the degree of precision, this inspection resolution is an improvement of more than 300% from a 150 ${\mu}m$ resolution, which is the conventional measurement criteria required for the deformation of neutron irradiated fuel rods.

• Nuclear Engineering and Technology
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• 제24권1호
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• pp.52-62
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• 1992

국내 원자력 발전소의 주요 기종인 가압 경수로에서는 일차 냉각계통을 통한 부식생성물 (CRUD)의 이동에 의해 노심에서 방사화된 후 노외표면에 침적된 방사성 핵종은 원전 종사자 방사선 피폭의 주원인이 된다. 따라서, 부식생성물에 의한 방사선 피폭을 감소시키기 위한 최적 화학운전 방안이 요망된다. 본 연구에서는 운전중인 국내 4개 발전소의 실제 수화학 운전 자료를 분석하였으며, 냉각재 화학운전 자료를 평가하기 위해 냉각재 수화학 조건에 따라 방사능 생성양을 계산할 수 있는 Computer 코드를 이용하였다. 실제 수화학 운전조건과 가정된 Elevated Li 운전조건에 따른 운전결과를 Computer 코드에 의해 예측하여 비교한 결과, Elevated Li 수화학 운전방법을 적용할 경우, 현재 적용되는 수화학 운전방법에 비하여 노심에서 부식생성물의 침적을 감소시킴으로써 노외 방사능 양을 상당히 감소시킬 수 있음을 알았다. 또한 계통 구성재 질과 핵연료봉의 건전성이 보장되는 한 냉각재 pH를 상승시키면 노외 방사능 생성감소에 유리함을 밝혔다.

• 한국가시화정보학회지
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• 제18권3호
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• pp.23-34
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• 2020

Two-phase flow and heat transfer characteristics during the reflood phase of a single heated rod in the KHU reflood experimental facility were examined. Two-phase flow behavior during the reflooding experiment was carefully visualized along with transient temperature measurement at a point inside the heated rod. By numerically solving one-dimensional inverse heat conduction equation using the measured temperature data, time-resolved wall heat flux and temperature histories at the interface of the heated rod and coolant were obtained. Once water coolant was injected into the test section from the bottom to reflood the heated rod of >700℃, vast vapor bubbles and droplets were generated near the reflood front and dispersed flow film boiling consisted of continuous vapor flow and tiny liquid droplets appeared in the upper part. Following the dispersed flow film boiling, inverted annular/slug/churn flow film boiling regimes were sequentially observed and the wall temperature gradually decreased. When so-called minimum film boiling temperature reached, the stable vapor film between the heated rod and coolant was suddenly collapsed, resulting in the quenching transition from film boiling into nucleate boiling. The moving speed of the quench front measured in the present study showed a good agreement with prediction by a correlation in literature. The obtained results revealed that typical two-phase flow and heat transfer behaviors during the reflood phase of overheated fuel rods in light water nuclear reactors are well reproduced in the KHU facility. Thus, the verified reflood experimental facility can be used to explore the effects of other affecting parameters, such as CRUD, on the reflood heat transfer behaviors in practical nuclear reactors.