• 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.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2002년도 학술대회지
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• pp.47-50
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• 2002

The numerical simulation of a 271-rod fuel assembly of nuclear Liquid-Metal Fast Reactor (LMFR) with an infernal blockage has been carried out. Internal blockage within a subassembly is addressed in the safety assessment because it potentially has very serious consequences for the reactor as a whole. Three dimensional calculations were performed using the SABRE4 computer code for the range of blockage positions and sizes to investigate the seriousness and detectability of the internal blockage. The magnitude and location of the peak temperatures together with the temperature distribution at the subassembly exit were calculated in order to look at the potential for damage within the subassembly, and the possibility of blockage detection. The analysis result shows that the 6-subchannel blockage causes large temperature rise within a assembly with practically no change in mixed mean temperature at the assembly exit.

• Nuclear Engineering and Technology
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• 제55권3호
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• pp.999-1008
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• 2023

Deformed fuel rods can cause a partial blockage of the flow area in a subchannel. Such flow blockage will influence the core coolant flow and further the core heat transfer during the reflooding phase and subsequent severe accidents. Nevertheless, most of the system analysis codes simulate the accident process based on the assumed flow blockage ratio, resulting in inconsistencies between simulated results and actual conditions. This paper aims to study the influence of flow blockage in severe accident scenario of the CAP1400 reactor. First, the flow blockage model of ISAA code is improved based on the FRTMB module. Then, the ISAA-FRTMB coupling system is adopted to model and calculate the QUENCH-LOCA-0 experiment. The correctness and validity of the flow blockage model are verified by comparing the peak cladding temperature. Finally, the DVI Line-SBLOCA accident is induced to analyze the influence of flow blockage on subsequent CAP1400 reactor core heat transfer and core degradation. From the results of the DVI Line-SBLOCA accident analysis, it can be concluded that the blockage ratio is in the range of 40%-60%, and the position of severe blockage is the same as that of cladding rupture. The blockage reduces the circulation area of the core coolant, which in turn impacts the heat exchange between the core and the coolant, leading to the early failure and collapse of some core assemblies and accelerating the core degradation process.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3154-3165
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• 2022

During a loss-of-coolant accident (LOCA) in the pressurized water reactor (PWR), there is a possibility that high temperature and internal pressure of the fuel rods lead to ballooning of the cladding, which causes a partial blockage of flow area in a subchannel. Such flow blockage would influence the core coolant flow, thus affecting the core heat transfer during a reflooding phase and subsequent severe accident. However, most of the system analysis codes simulate the accident process based on the assumed channel blockage ratio, resulting in the fact that the simulation results are not consistent with the actual situation. This paper integrates the developed core Fuel Rod Thermal-Mechanical Behavior analysis (FRTMB) module into the self-developed severe accident analysis code ISAA. At the same time, the existing flow blockage model is improved to make it possible to simulate the change of flow distribution due to fuel rod deformation. Finally, the ISAA-FRTMB is used to simulate the QUENCH-LOCA-0 experiment to verify the correctness and effectiveness of the improved flow blockage model, and then the effect of clad ballooning on core heat transfer and subsequent parts of core degradation is analyzed.

• 한국통신학회논문지
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• 제39C권9호
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• pp.757-766
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• 2014

본 논문에서는 차기 군 위성통신체계 환경에서 OTM(On-The-Move) 위성단말의 채널 blockage 신호 손실 보상을 위한 협업통신 기법을 제안하였다. 제안하는 협업기법은 인접 OTM 위성단말과 지상 무전기 체계를 통해 데이터를 공유하고 직교확산코드를 이용하여 대역확산 후 동시에 전송한다. 중첩되어 전송된 확산열은 EGC(Equal Gain Combining) 방식으로 결합한다. 성능 분석을 위해 OTM 위성단말의 blockage 채널을 2-state Markov chain으로 모델링하였으며, 이를 기반으로 협업단말들의 blockage 채널 상태에 따른 비트오류율을 도출하였다. 성능분석 결과 채널 조건이 더 나은 인접 OTM 위성단말의 협업으로 비트오류율 성능이 향상됨을 확인할 수 있었다. 특히 blockage 확률이 높을수록 협업을 통해 더욱 우수한 성능을 확인할 수 있었다. 그러나 서로 다른 지연을 갖는 확산 열 간 중첩으로 인해 협업단말 수 증가 시 다중접속간섭으로 인한 성능상의 제약이 있음을 확인하였다.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3217-3228
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• 2021

Flow blockage of the fuel assembly in the lead-based fast reactor (LFR) may produce critical local spots, which will result in cladding failure and threaten reactor safety. In this study, the flow blockage characteristics were analyzed with the sub-channel analysis method, and the circumferentially-varied method was employed for considering the non-uniform distribution of circumferential temperature. The developed sub-channel analysis code SACOS-PB was validated by a heat transfer experiment in a blocked 19-rod bundle cooled by lead-bismuth eutectic. The deviations between the predicted coolant temperature and experimental values are within ±5%, including small and large flow blockage scenarios. And the temperature distributions of the fuel rod could be better simulated by the circumferentially-varied method for the small blockage scenario. Based on the validated code, the analysis of blockage characteristics was conducted. It could be seen from the temperature and flow distributions that a large blockage accident is more destructive compared with a small one. The sensitivity analysis shows that the closer the blockage location is to the exit, the more dangerous the accident is. Similarly, a larger blockage length will lead to a more serious case. And a higher exit temperature will be generated resulting from a higher peak coolant temperature of the blocked region. This work could provide a reference for the future design and development of the LFR.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2011년 춘계학술대회논문집
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• pp.269-274
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• 2011

A CFD analysis was performed to examine the inner channel blockage of dual-cooled fuel which has being developed for the power uprate of a pressurized water reactor (PWR). The dual-cooled fuel consists of an annular fuel pellet($UO_2$) and dual claddings as well as internal and external cooling channels. The dual-cooled annular fuel is different from a conventional solid 려el by employing an internal cooling channel for each fuel pellet as well as an external cooling channel. One of the key issues is the hypothetical event of inner channel blockage because the inner channel is an isolated flow channel without the coolant mixing between the neighboring flow channels. The inner channel blockage could cause the Departure from Nucleate Boiling (DNB) in the inner channel that eventually causes a fuel failure. This paper presents the CFD simulation of the flow through the side holes of the bottom end plug for the complete entrance blockage of the inner channel. Since the amount of coolant supply to the inner channel depends on largely the pressure loss at the side hole, the pressure loss coefficient of the side hole was estimated by the CFD analysis. The CFD prediction of the loss coefficient showed a reasonable agreement with an experimental data for the complete blockage of both the inner channel entrance and the outer channel. The CFD predictions also showed the decrease of the loss coefficient as the outer channel blockage increases.

• 대한기계학회논문집
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• 제6권3호
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• pp.211-221
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• 1982

Numerical analysis has been made on the heat transfer of two dimensional turbulent channel with a slat type blockage. Especially the effects of the height of slat and Reynolds number on the heat transfer characteristics of channel wall have been investigated. The methods of accelerating the convergence of the numerical solution of governing differential equation have been also examined. Line-by-line iterative method shows higher convergence rate than point-by-point iterative method for solution of both momentum equation and energy equation. The results show that the ratio of heat transfer coefficient of the wall near the blockage to that of the fully developed flow increase with increasing the ratio of blockage to channel height and decreasing the Reynolds number. These trends of variation of heat transfer coefficient with respect to the height of slat and Reynolds number agree with those of Sparrow's experiment on the pipe flow with slat type blockage.

• Nuclear Engineering and Technology
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• 제41권6호
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• pp.797-806
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• 2009

The Multichannel Analyzer for Transient and steady-state in Rod Array - Liquid Metal Reactor for Flow Blockage analysis (MATRA-LMR-FB) code for the analysis of a subchannel blockage has been developed and evaluated through several experiments. The current version of the code is improved here by the implementation of a distributed resistance model which accurately considers the effect of flow resistance on wire spacers, by the addition of a turbulent mixing model, and by the application of a hybrid scheme for low flow regions. Validation calculations for the MATRA-LMR-FB code were performed for Oak Ridge National Laboratory (ORNL) 19-pin tests with wire spacers and Karlsruhe 169-pin tests with grid spacers. The analysis of the ORNL 19-pin tests conducted using the code reveals that the code has sufficient predictive accuracy, within a range of 5 $^{\circ}C$, for the experimental data with a blockage. As for the results of the analyses, the standard deviation for the Karlsruhe 169-pin tests, 0.316, was larger than the standard deviation for the ORNL 19-pin tests, 0.047.

• 대한조선학회논문집
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• 제43권4호
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• pp.431-439
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• 2006

Due to the limitation of size of the test section, blockage effects could not be avoided in the model test of yacht sails for common wind tunnels. In this paper, a numerical analysis is performed to investigate the blockage effects on the lift and drag forces measured from wind tunnel experiments for a 30 feet sloop yacht sail. Complex airflows around the jib and main sails including three-dimensional flow separations are calculated for various close-hauled conditions. It is found that the blockage of a wind tunnel changes the flow separation and consequently the lift and drag forces of the sails, especially the main sail, reduce and increase, respectively, due to the blockage effects.