• Nuclear Engineering and Technology
• /
• 제53권9호
• /
• pp.2847-2858
• /
• 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
• /
• 제55권3호
• /
• pp.999-1008
• /
• 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
• /
• 제39권1호
• /
• pp.43-50
• /
• 2007

A partial flow blockage in an assembly of a liquid metal reactor could result in a cooling deficiency of the core. To develop a partial blockage detection system, we have studied the changes of the temperature fluctuation characteristics in the upper plenum according to changes of the t10w blockage conditions in an assembly. We analyzed the temperature fluctuation in the upper plenum with the Large Eddy Simulation (LES) turbulence model in the CFX code and evaluated its statistical parameters. Based on the results of the statistical analyses, we developed a neural network model for detecting a partial flow blockage in an assembly. The neural network model can retrieve the size and the location of a flow blockage in an assembly from a change of the root mean square, the standard deviation, and the skewness in the temperature fluctuation data. The neural network model was found to be a possible alternative by which to identify a flow blockage in an assembly of a liquid metal reactor through learning and validating various flow blockage conditions.

• Wind and Structures
• /
• 제12권6호
• /
• pp.541-551
• /
• 2009

This paper explains how to correctly measure the drag coefficient of a circular cylinder in wind tunnels with large blockage ratios and for the sub-critical to the super-critical flow regimes. When dealing with large blockage ratios, the drag has to be corrected for wall constraints. Different formulations for correcting blockage effect are compared for each flow regime based on drag measurements of smooth circular cylinders performed in a wind tunnel for three different blockage ratios. None of the correction model known in the literature is valid for all the flow regimes. To optimize the correction and reduce the scatter of the results, different correction models should be combined depending on the flow regime. In the sub-critical regime, the best results are obtained using Allen and Vincenti's formula or Maskell's theory with ${\varepsilon}$=0.96. In the super-critical regime, one should prefer using Glauert's formula with G=0.6 or the model of Modi and El-Sherbiny. The change in the formulations appears at the flow transition with a variation of the wake pattern when passing from sub-critical to super-critical flow regimes. This parameter being not considered in the known blockage corrections, these theories are not valid for all the flow regimes.

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

• Wind and Structures
• /
• 제30권6호
• /
• pp.597-616
• /
• 2020

The blockage effect on the aerodynamic characteristics of tall buildings is a fundamental issue in wind tunnel test but has rarely been addressed. To evaluate the blockage effects on the aerodynamic forces on a square tall building and flow field peripherally, large eddy simulations (LES) were performed on a 3D square cylinder with an aspect ratio of 6:1 under the uniform smooth inflow and turbulent atmospheric boundary layer (ABL) inflow generated by the narrowband synthesis random flow generator (NSRFG). First, a basic case at a blockage ratio (BR) of 0.8% was conducted to validate the adopted numerical methodology. Subsequently, simulations were systematically performed at 6 different BRs. The simulation results were compared in detail to illustrate the differences induced by the blockage, and the mechanism of the blockage effects under turbulent inflow was emphatically analysed. The results reveal that the pressure coefficients, the aerodynamic forces, and the Strouhal number increase monotonically with BRs. Additionally, the increase of BR leads to more coherence of the turbulent structures and the higher intensity of the vortices in the vicinity of the building. Moreover, the blockage effects on the aerodynamic forces and flow field are more significant under smooth inflow than those under turbulent inflow.

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

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

• Wind and Structures
• /
• 제12권1호
• /
• pp.49-61
• /
• 2009

A method is presented to estimate the form drag and the base pressure on a triangular cylinder in the presence of blockage effect. The Strouhal number, which is found to increase with the flow constriction experimentally by Ramamurthy & Ng (1973), may be decoupled from the blockage effect when re-defined by using the velocity at flow separation and a theoretical wake width. By incorporating this wake width into the momentum equation by Maskell (1963) for the confined flow, a relationship between the form drag and the base pressure is derived. Independently, the experimental data of surface pressure from Ramamurthy & Lee (1973) are found to be independent of the blockage effect when expressed in terms of a modified pressure coefficient involving the pressure at separation. Using the potential flow model by Parkinson & Jandali (1970) and its subsequent development in Yeung & Parkinson (2000) for the unconfined flow, a linear relation between the pressure at separation and the form drag is formulated. By solving the two equations simultaneously with a specified blockage ratio and an apex angle of the triangular cylinder, the predictions of the drag and the base pressure are in reasonable agreement with experimental data. A new theoretical relationship for the Strouhal number, pressure drag coefficient and base pressure proposed in this study allows the confinement effect to be appropriately taken into consideration. The present approach may be extended to three-dimensional bluff bodies.

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