• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1174-1183
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• 2016

Hydrogen release during severe accidents poses a serious threat to containment integrity. Mitigating procedures are necessary to prevent global or local explosions, especially in large steel shell containments. The management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity. The MELCOR 1.8.6 was used to assess core degradation and containment behavior during the large break loss-of-coolant accident without the actuation of the safety injection system except for accumulators in Beznau nuclear power plant. Also, hydrogen distribution in containment and performance of hydrogen reduction system were investigated.

• International Journal of Air-Conditioning and Refrigeration
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• 제8권2호
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• pp.51-59
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• 2000

The purpose of the present study is to investigate the absorbing characteristics in a vertical falling film type absorber using LiBr-H$H_2O$ solution as working fluids with the concentration of 60 wt%. The experimental apparatus consists of an absorber with the diameter of 17.2 mm and the length of 1150 mm, a generator, an evaporator (condenser), a weak solution tank and a sampling trap device and so on. The parameters were the solution temperatures of 45 and 50$^{\circ}$C, coolant temperatures of 30 and 35$^{\circ}$C, and the film Reynolds numbers from 50 to 150. The pressure drop in the absorber increased as the solution and coolant temperatures decreased. The pressure drop in the absorber increased up to the film Reynolds number of 90, however, decreased at the film Reynolds number above 90. The maximum absorption mass flux was observed at the film Reynolds number of 90. Absorption mass fluxes increased as the coolant temperature decreased. Accordingly, absorption mass fluxes and heat transfer coefficients under the subcooled condition increased more than those under the superheated condition. It is claimed that heat transfer coefficients are deeply affected by the solution temperature more than the coolant temperature within the experimental range.

• 한국추진공학회:학술대회논문집
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• 한국추진공학회 2017년도 제48회 춘계학술대회논문집
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• pp.591-595
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• 2017

이상유동 모델과 동적격자계를 활용하여 탄의 사출관 내부의 열유동과 탄의 운동성능을 해석하는 수치모사를 진행하였다. 고온의 공기와 냉각제간의 상호작용 및 유동장을 해석하기 위해, Realizable $k-{\varepsilon}$ 난류 모델과 VOF (Volume Of Fluid) 모델을 선정하고 냉각제 유량변이에 따른 수치 해석을 진행하였다. 해석결과, 사출관의 압력은 냉각제의 유뮤에 따라 큰 차이를 보였고, 냉각제량에 따라서도 각각의 차이를 보였다. 탄의 속도와 가속도의 변이는 압력에 종속하여 나타났다.

• 한국추진공학회:학술대회논문집
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• 한국추진공학회 2017년도 제48회 춘계학술대회논문집
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• pp.1021-1025
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• 2017

열교환기 지상시험 설비의 내구성 확보를 위해 필수적인 냉각수 매니폴드에 대해 열/유동해석을 진행했으며, 분사기와 유로의 배열 등의 형상을 결정해 개발 중인 엔진의 헤드에 적용하였다. 제작된 엔진 헤드에 대한 검증시험이 진행됐으며, 엔진의 분사기면에 도포된 열차단코팅(TBC) 등에서 열적 손상이 확인되지 않았다. 연소시험 결과와 수치해석을 비교하면 냉각수 출구온도가 $15^{\circ}C$ 정도의 차이를 보이지만 냉각수 매니폴드 상부에 위치하는 액체산소 매니폴드, 열 차폐코팅, 화염면의 위치 등을 감안하면 합당한 수준으로 판단된다.

• Nuclear Engineering and Technology
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• 제26권2호
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• pp.161-167
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• 1994

본 논문은 비상노심냉각계통을 필요로 하지 않는 압력관형 피동 수냉각로 개념을 제시한다. 여기서는 사고시 핵연료에서 생성되는 열을 감속재로 효과적으로 전달시키기 위해 금속 핵연료 매트릭스를 사용하는 핵연료 채널을 채택한다. 정상 운전시에는 보통의 냉각재가 핵연료를 냉각시키지만, 냉각재상실사고를 포함하여 정상적인 냉각계통의 작동이 이루어지지 않을 경우에는 피동 감속재냉각계통에 의해 핵연료가 냉각된다. 유한요소 코드를 이용한 해석 결과, 정상 상태 및 사고시 핵연료 온도를 허용 한도 이내로 유지할 수 있는 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제41권9호
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• pp.1215-1222
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• 2009

The present work pertains to a research program to study Molten Fuel-Coolant Interactions (MFCI), which may occur in a nuclear power plant during a hypothetical severe accident. Dynamics of the hot liquid (melt) droplet and the volatile liquid (coolant) were investigated in the MISTEE (Micro-Interactions in Steam Explosion Experiments) facility by performing well-controlled, externally triggered, single-droplet experiments, using a high-speed visualization system with synchronized digital cinematography and continuous X-ray radiography. The current study is concerned with the MISTEE-NCG test campaign, in which a considerable amount of non-condensable gases (NCG) are present in the film that enfolds the molten droplet. The SHARP images for the MISTEE-NCG tests were analyzed and special attention was given to the morphology (aspect ratio) and dynamics of the air/ vapor bubble, as well as the melt drop preconditioning. Energetics of the vapor explosion (conversion ratio) were also evaluated. The MISTEE-NCG tests showed two main aspects when compared to the MISTEE test series (without entrapped air). First, analysis showed that the melt preconditioning still strongly depends on the coolant subcooling. Second, in respect to the energetics, the tests consistently showed a reduced conversion ratio compared to that of the MISTEE test series.

• Nuclear Engineering and Technology
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• 제49권3호
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• pp.466-475
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• 2017

Coolant mixing under natural circulation flow regime constitutes a key parameter that may play a role in the course of an accidental transient in a nuclear pressurized water reactor. This issue has motivated some experimental investigations carried out within the OECD/NEA PKL projects. The aim was to assess the coolant mixing phenomenon in the reactor pressure vessel downcomer and the core lower plenum under several asymmetric steady and unsteady flow conditions, and to provide experimental data for code validations. Former studies addressed the mixing phenomenon using, on the one hand, one-dimensional computational approaches with cross flows that are not fully validated under transient conditions and, on the other hand, expensive computational fluid dynamic tools that are not always justified for large-scale macroscopic phenomena. In the current framework, an unsteady coolant mixing experiment carried out in the Rossendorf coolant mixing test facility is simulated using the three-dimensional porous media capabilities of the thermal-hydraulic system CATHARE code. The current study allows highlighting the current capabilities of these codes and their suitability for reproducing the main phenomena occurring during asymmetric transient natural circulation mixing conditions.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.968-978
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• 2017

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal-hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2847-2858
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• 2021

When a nuclear reactor with rectangular fuel assemblies runs for a long time, impurities and debris may be taken into coolant channels, which may cause flow blockage, and the blocked fuel assemblies might be destroyed. Therefore, the purpose of this study is to perform a thermal-hydraulic analysis of a rectangular fuel assembly by STAR-CCM+, under the condition of one subchannel with 80% blockage ratio. A rectangular fuel assembly of the International Atomic Energy Agency (IAEA) 10 MW material test reactor (MTR) is chosen. In view of the gasket material taken into the coolant channel is close to the single side of the coolant channel, in the flow blockage accident of the Oak Ridge Research Reactor (ORRR), a new blockage category called single side blockage is attempted. The blockage positions include inlet, middle and outlet, and the blockage is set as a cuboid. It is found by simulations that the blockage redistributes the mass flow rate, and large vortices appear locally. The peak temperature of the cladding is maximum, when the blockage is located at the single side of the coolant channel inlet, and no boiling occurs in all blockage cases. Moreover, as the height of the blockage increases, the damage caused by the blockage increases slightly.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3182-3195
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• 2021

The implementation and validation of multi-dimensional (multi-D) features in thermal-hydraulic system codes aims to extend the application of these codes towards multi-scale simulations. The main goal is the simulation of large-scale three-dimensional effects inside large volumes such as piping or vessel. This novel approach becomes especially relevant during the simulation of accidents with strongly asymmetric flow conditions entailing density gradients. Under such conditions, coolant mixing is a key phenomenon on the eventual variation of the coolant temperature and/or boron concentration at the core inlet and on the extent of a local re-criticality based on the reactivity feedback effects. This approach presents several advantages compared to CFD calculations, mainly concerning the model size and computational efforts. However, the range of applicability and accuracy of the newly implemented physical models at this point is still limited and needs to be further extended. This paper aims at contributing to the validation of the multi-D features of the system code ATHLET based on the simulation of the Tests 1.1 and 2.1, conducted at the test facility ROCOM. Overall, the multi-D features of ATHLET predict reasonably well the evolution from both experiments, despite an observed overprediction of coolant mixing at the vessel during both experiments.