• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.391-396
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• 1996

내부부수로, 벽면부수로, 모서리부수로를 포함하는 가압경수로형 원자로의 핵연료집합체를 모의하는 3$\times$3 봉다발을 모델로 수치해석을 통해 봉다발 주변의 유동특성을 알아보고 각 봉에서의 원주방향 위치에 따른 국소열전달 특성에 관해 고찰하였다. 봉다발에서 열전달계수의 분포는 벽면영향으로 인한 각 부수로에서의 유속분포와 밀접한 관계가 있으며 내부부수로에 인접한 봉에서 가장 높았고, 그 다음이 벽면부수로, 모서리부수로에 인접한 봉에서는 가장 작게 나타났다. 현재 핵연료의 열수력 설계시에 적용하고 있는 부수고 내의 모든 열수력학적 변수가 일정하다고 가정하는 부수로 해석방법은 봉다발내의 실제 열전달 현상과는 상당한 차이가 있음을 보여주었다.

• Journal of the Korean Society of Radiology
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• v.11 no.7
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• pp.571-578
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• 2017

Bubbles are generated by the boiling of the cooling water when an accident occurs in the reactor and then in order to measure the void fraction, the Optical Fiber Probe(OFP) and optical camera are used in thermal hydraulic safety research. However, such an optical method is not suitable for measuring the void fraction in a $17{\times}17$ array of fuel rods due to the geometrical limitations. This study was conducted as a preliminary study using x-ray system and various phantoms before applying to rod bundles. Through radiographic and tomographic experiments, the tube voltage of the x-ray generator was 130 kVp and the tube current was 1 mA. In addition, it is possible to measure the hole of 1mm in size visually through the bubble resolution phantom, and it is confirmed that the contrast is relatively decreased in the inside of the freon in the case of the contrast evaluation using the road phantom. However, we could obtain good image without distortion when reconstructing the image. Bubble generation phantom experiments were used to confirm the flow direction of the bubbles and to acquire tomography images. The image J tool was used to measure the void fraction of 18 % for a single tomography image. This study has carried out previous researches for the measurement of the bubble rate around the nuclear fuel and could be used as a basic research for continuous research.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.40 no.3
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• pp.165-172
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• 2016

An experimental study of the thermal-hydraulic characteristics of passive safety systems (PSSs) was conducted using a system-integrated modular advanced reactor-integral test loop (SMART-ITL). The present passive safety injection system for the SMART-ITL consists of one train with the core makeup tank (CMT), the safety injection tank, and the automatic depressurization system. The objective of this study is to investigate the injection effect of the PSS on the small-break loss-of-coolant accident (SBLOCA) scenario for a 0.4 inch line break in the safety-injection system (SIS). The steady-state condition was maintained for 746 seconds before the break. When the major parameters of the target value and test results were compared, most of the thermal-hydraulic parameters agreed closely with each other. The water level of the reactor pressure vessel (RPV) was maintained higher than that of the fuel assembly plate during the transient, for the present CMT and safety injection tank (SIT) flow rate conditions. It can be seen that the capability of an emergency core cooling system is sufficient during the transient with SMART passive SISs.