• 에너지공학
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• 제24권1호
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• pp.42-50
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• 2015

신형 원전의 피동격납건물냉각계통(PCCS: Passive Containment Cooling System)을 구성하는 단일 전열관의 열제거 성능을 평가하기 위해, 비응축성 기체 존재 시 수직 튜브 외벽에서 발생하는 증기의 응축 열전달에 대한 실험을 수행하였다. 외경 40 mm, 길이 1.0 m의 전열관 외벽에서 증기-공기 혼합물의 평균 열전달계수를 측정하였으며, 압력 2-4 bar, 공기의 질량분율 0.1-0.7의 범위에서 실험데이터를 수집하였다. 이를 통해 압력과 비응축성기체의 농도가 응축 열전달계수에 미치는 영향을 평가하였다. 실험결과를 기존의 열전달모델인 Uchida와 Dehbi의 상관식과 비교하였으며, 이들 상관식은 실험결과에 비해 상대적으로 열전달계수를 낮게 예측함을 확인하였다.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.2997-3009
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• 2023

The Passive containment Air-cooling System (PAS) can effectively remove the decay heat of the modular small nuclear reactor after an accident. The details of natural convection around the dome, which is a key part of PAS, were investigated numerically in the present study. The thermal dynamics around the dome were studied through the temperature, pressure and velocity contours and the streamlines. Additionally, the formation of the buoyant plume at the top of the dome was investigated. The results show that with the increase of Ra, the lift-off point moves toward the bottom of the dome, and the eddy under the buoyant plume grows larger gradually, which enhances the heat transfer. And the heat transfer along the dome surface with different truncation angles was investigated. As the angle increases, the heat transfer coefficient becomes stronger as well. Consequently, a newly developed heat transfer correlation considering the influence of truncation angle for the dome is proposed based on the simulated results. This study could provide a better understanding of natural convection around the dome of PAS and the proposed correlation could also offer more predictive value in the improvement of nuclear safety.

• Nuclear Engineering and Technology
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• 제41권8호
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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.

• 에너지공학
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• 제7권1호
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• pp.122-130
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• 1998

A simplified heat transfer model for the cooling capability of the AP 600 PCCS is proposed I this paper. As the PCCS domain is covered with very thin and long water film, it is phenomenologically divided into 3 regions; water entrance effect region, asymptotic region, and air entrance effect region. As the length of the asymptotic region is estimated to be over 90% of the whole domain, the phenomena in the asymptotic region is focused. Using the analogy between heat and mass transfer phenomena in a turbulent situation, a new dependent variable combining temperature and vapor mass fraction was defined. The similarity between the PCCs phenomena in the asymptotic region and the buoyant air flow phenomena on a vertical heated plate is derived. Using the similarity, the simplified heat transfer correlations for the interfacial heat fluxes and the ratios of latent heat transfer to sensible heat transfer were established. To verify the accuracy of the correlation, the results of this study were compared with those of other numerical analyses performed for the same configuration and they are well within the range of 15% difference.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.605-612
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• 2013

Removal of decay heat from an operating reactor during a prolonged station blackout condition is a big concern for reactor designers, especially after the recent Fukushima accident. In the case of a prolonged station blackout condition, heat removal is possible only by passive means since no pumps or active systems are available. Keeping this in mind, the AHWR has been designed with many passive safety features. One of them is a passive means of removing decay heat with the help of Isolation Condensers (ICs) which are submerged in a big water pool called the Gravity Driven Water Pool (GDWP). The ICs have many tubes in which the steam, generated by the reactor core due to the decay heat, flows and condenses by rejecting the heat into the water pool. After condensation, the condensate falls back into the steam drum of the reactor. The GDWP tank holds a large amount of water, about 8000 $m^3$, which is located at a higher elevation than the steam drum of the reactor in order to promote natural circulation. Due to the recent Fukushima type accidents, it has been a concern to understand and evaluate the capability of the ICs to remove decay heat for a prolonged period without escalating fuel sheath temperature. In view of this, an analysis has been performed for decay heat removal characteristics over several days of an AHWR by ICs. The computer code RELAP5/MOD3.2 was used for this purpose. Results indicate that the ICs can remove the decay heat for more than 10 days without causing any bulk boiling in the GDWP. After that, decay heat can be removed for more than 40 days by boiling off the pool inventory. The pressure inside the containment does not exceed the design pressure even after 10 days by condensation of steam generated from the GDWP on the walls of containment and on the Passive Containment Cooling System (PCCS) tubes. If venting is carried out after this period, the decay heat can be removed for more than 50 days without exceeding the design limits.

• 에너지공학
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• 제27권4호
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• pp.27-35
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• 2018

피동원자로건물냉각계통(PCCS)은 사고 발생 시 원자로건물로 방출된 열을 제거하여 원전의 건전성을 보장하기 위해 설계되었다. PCCS의 열제거 성능은 증기-공기 혼합물의 응축열전달에 의해 결정된다. 본 연구에서는 응축열전달계수의 예측 정확도를 향상시키기 위해 새로운 상관식을 이식한 MARS-KS 코드를 사용하여 PCCS의 열제거 성능을 평가하였다. MARS-KS 코드에 사용된 새로운 상관식은 압력, 벽면과냉도, 비응축성 기체 질량분율 및 응축튜브의 종횡비와 같은 열전달계수에 영향을 미치는 변수들을 이용하여 개발하였고, 이는 MARS-KS코드의 기본 응축 모델인 Colburn-Hougen 모델을 대체하여 적용되었다. 대형파단 냉각재상실사고 발생 시 PCCS의 운전에 따른 다양한 열수력학적 변수들을 분석하였고, 열제거 성능 평가를 위해 새로운 상관식이 적용된 MARS-KS 코드의 원자로건물 압력거동 계산결과와 기존의 응축모델을 이용한 해석결과를 비교하였다.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.949-958
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• 2024

In contrast to all-time flooded small modular reactor (SMR) systems, an in-kind flooding safety system (FSS) has been proposed as a passive safety system applicable to small modular reactors (SMRs) that adopt a metal containment vessel (MCV). Under transient conditions, the FSS can provide emergency cooling to dry reactor cavities and sustain long-term coolability using re-acquired evaporated steam in the reactor building on demand. When designing an FSS, the effect of the flooding flow area is vital as it affects the overall accident sequence and safety. Therefore, in this study, a MELCOR model of a reference SMR is developed and numerical analysis is performed under postulated accident scenarios. Without flooding, the MCV pressure of the reactor module exceeds the design pressure before core damage. To prevent core damage, an emergency flooding strategy is devised using various flow path parameters and requirements to ensure an adequate emergency coolant supply before the core damage is investigated. The results indicate that a flow area exceeding 0.02 m2 is required in the FSS to prevent MCV overpressure and core damage. This study is the first to report a strategic analysis for appropriately sizing an FSS flooding valve applicable to innovative SMRs.

• 에너지공학
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• 제25권1호
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• pp.56-68
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• 2016

최근 원자력 발전소의 안전성을 획기적으로 향상시키기 위한 연구가 활발하게 진행되고 있으며 특히 피동냉각계통의 연구개발이 아주 중요하게 부각되고 있다. 피동냉각계통의 열전달 방식으로는 응축열전달 양식이 주로 채택되고 있다. 이와 같은 맥락에서 부산대학교 Ahn & Yun (Ahn 등, 2014)은 새로운 수평관내부 응축 모델을 제시한 바 있다. 본 연구에서는 먼저 Ahn & Yun 이 제시한 수평관 응축 모델을 MARS 코드에 삽입하고 PASCAL 실험데이터를 이용하여 평가하였다. 이 평가결과를 통해 Ahn & Yun 모델의 코드적용에 있어 문제점을 규명하고 새로운 적용방법론을 적용하여 다양한 실험데이터로 다시 평가함으로써 MARS 코드의 향상된 응축 열전달 해석 능력을 확인하였다.

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4329-4337
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• 2022

Jet impinging device is designed for decay heat removal on horizontal fuel rods in a low temperature heating reactor. An experimental system with a fuel rod simulator is established and experiments are performed to evaluate water film covering capacity, within 0.0287-0.0444 kg/ms mass flow rate, 0-164.1 kW/m² heating flux and 13.8-91.4℃ feeding water temperature. An effective method to obtain the film coverage rate by infrared equipment is proposed. Water film flowing patterns are recoded and the film coverage rates at different circumference angles are measured. It is found the film coverage rate decreases with heating flux during single-phase convection, while increases after onset of nucleate boiling. Besides, film coverage rate is found affected by Marangoni effect and film accelerating effect, and surface wetting is significantly facilitated by bubble behavior. Based on the observed phenomenon and physical mechanism, dry-out depth and initial dry-out rate are proposed to evaluate film covering potential on a heating surface. A model to predict film coverage rate is proposed based on the data. The findings would have reliable guide and important implications for further evaluation and design of decay heat removal system of new reactors, and could be helpful for passive containment cooling research.