• Nuclear Engineering and Technology
• /
• v.51 no.6
• /
• pp.1532-1539
• /
• 2019

This paper analyzes a new method for reducing boiling water reactor fuel temperature during a Loss of Coolant Accident (LOCA). The method uses a device called Reverse Flow Restriction Device (RFRD) at the inlet of fuel bundles in the core to prevent coolant loss from the bundle inlet due to the reverse flow after a large break in the recirculation loop. The device allows for flow in the forward direction which occurs during normal operation, while after the break, the RFRD device changes its status to prevent reverse flow. In this paper, a detailed simulation of LOCA has been carried out using the U.S. NRC's TRACE code to investigate the effect of RFRD on the flow rate as well as peak clad temperature of BWR fuel bundles during three different LOCA scenarios: small break LOCA (25% LOCA), large break LOCA (100% LOCA), and double-ended guillotine break (200% LOCA). The results demonstrated that the device could substantially block flow reversal in fuel bundles during LOCA, allowing for coolant to remain in the core during the coolant blowdown phase. The device can retain additional cooling water after activating the emergency systems, which maintains the peak clad temperature at lower levels. Moreover, the RFRD achieved the reflood phase (when the saturation temperature of the clad is restored) earlier than without the RFRD.

• 한국연소학회:학술대회논문집
• /
• 2014.11a
• /
• pp.165-166
• /
• 2014

The application of particle reaction model in the packed bed process modeling is discussed for iron ore pellet induration process. Combustion of coke breeze in the pellet is estimated by using shrinking unreacted-core model and grain model in which the progress of chemical reaction is described in different concepts. Under the identical inlet gas and solid conditions, the calculation using shrinking core model showed deviated results in terms of temperature profile and conversion fraction, which may imply the significance of selecting proper particle reaction model in consideration of particle characteristics and process operation conditions.

• The KSFM Journal of Fluid Machinery
• /
• v.15 no.4
• /
• pp.5-11
• /
• 2012

To identify the flow characteristics of the SMART reactor, a flow distribution model test and a numerical simulation are performed in KAERI. Among several part of the SMART reactor, the fuel assemblies are simulated using simulators because of the complexity. The geometries of the core in the SMART reactor and simulator are different, but some similarities are maintained such as the ratio of pressure drop in the vertical and cross directions. There are cross flow holes in each core simulator to reproduce the cross flow of SMART fuel assemblies. To know the flow characteristics of the cross flow, numerical analysis is performed. As the cross flow area is decreased, the pressure drop between inlet and outlet is decreased. Also, when the flow imbalance between two core simulators is constant, the cross flow area does not significantly affect the cross flow.

• Nuclear Engineering and Technology
• /
• v.44 no.7
• /
• pp.735-744
• /
• 2012

In order to quantify the flow distribution characteristics of APR+ reactor, a test was performed on a test facility, ACOP ($\underline{A}$PR+ $\underline{C}$ore Flow & $\underline{P}$ressure Test Facility), having a length scale of 1/5 referring to the prototype plant. The major parameters are core inlet flow and outlet pressure distribution and sectional pressure drops along the major flow path inside reactor vessel. To preserve the flow characteristics of prototype plant, the test facility was designed based on a preservation of major flow path geometry. An Euler number is considered as primary dimensionless parameter, which is conserved with a 1/40.9 of Reynolds number scaling ratio. ACOP simplifies each fuel assembly into a hydraulic simulator having the same axial flow resistance and lateral cross flow characteristics. In order to supply boundary condition to estimate thermal margins of the reactor, the distribution of inlet core flow and core exit pressure were measured in each of 257 fuel assembly simulators. In total, 584 points of static pressure and differential pressures were measured with a limited number of differential pressure transmitters by developing a sequential operation system of valves. In the current study, reactor flow characteristics under the balanced four-cold leg flow conditions at each of the cold legs were quantified, which is a part of the test matrix composing the APR+ flow distribution test program. The final identification of the reactor flow distribution was obtained by ensemble averaging 15 independent test data. The details of the design of the test facility, experiment, and data analysis are included in the current paper.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.37 no.9
• /
• pp.855-862
• /
• 2013

A series of 1/5 scale-down reactor flow distribution tests had been conducted to determine the hydraulic characteristics of an APR+ (Advanced Power Reactor Plus), which were used as the input data for an open core thermal margin analysis code. In this study, to examine the applicability of computational fluid dynamics with the porous model to the analysis of APR+ internal flow, simulations were conducted using the commercial multi-purpose computational fluid dynamics software ANSYS CFX V.14. It was concluded that the porous domain approach for some reactor internal structures could adequately predict the flow characteristics inside a reactor in a qualitative manner. If sufficient computational resources are available, the predicted core inlet flow distribution is expected to be more accurate by considering the real geometry of the internal structures, especially upstream of the core inlet.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
• /
• v.29 no.12
• /
• pp.654-662
• /
• 2017

When emergency core cooling system (ECCS) is operated during loss of coolant accident (LOCA) in a pressurized water reactor (PWR), pressurized thermal shock (PTS) phenomenon can occur as cooling water is injected into a cold leg, mixed with hot primary coolant, and then entrained into a reactor vessel. Insufficient flow mixing may cause temperature stratification and steam condensation. In addition, flow vibration may cause thermal stresses in surrounding structures. This will reduce the life of the reactor vessel. Due to the importance of PTS phenomenon, in this study, calculation was performed for Test 1 among six types of OECD/NEA ROSA tests with ANSYS CFX R.17. Predicted results were then compared to measured data. Additionally, because temperature difference between the hot coolant at the inlet of the cold leg and the cold cooling water at the inlet of the ECCS injection line is 200 K or more, buoyancy force due to density difference might have significant effect on thermal-hydraulic characteristics of flow. Therefore, in this study, the necessity to include buoyancy force term in governing equations for accurate prediction of single phase thermal stratification in both cold legs and downcomer by ECCS injection was numerically studied.

• Nuclear Engineering and Technology
• /
• v.53 no.10
• /
• pp.3275-3285
• /
• 2021

A Nuclear Power Plant (NPP) is a complex dynamic system-of-systems with highly nonlinear behaviors. In order to control the plant operation under both normal and abnormal conditions, the different systems in NPPs (e.g., the reactor core components, primary and secondary coolant systems) are usually monitored continuously, resulting in very large amounts of data. This situation makes it possible to integrate relevant qualitative and quantitative knowledge with artificial intelligence techniques to provide faster and more accurate behavior predictions, leading to more rapid decisions, based on actual NPP operation data. Data-driven models (DDM) rely on artificial intelligence to learn autonomously based on patterns in data, and they represent alternatives to physics-based models that typically require significant computational resources and might not fully represent the actual operation conditions of an NPP. In this study, a feed-forward backpropagation artificial neural network (ANN) model was trained to simulate the interaction between the reactor core and the primary and secondary coolant systems in a pressurized water reactor. The transients used for model training included perturbations in reactivity, steam valve coefficient, reactor core inlet temperature, and steam generator inlet temperature. Uncertainties of the plant physical parameters and operating conditions were also incorporated in these transients. Eight training functions were adopted during the training stage to develop the most efficient network. The developed ANN model predictions were subsequently tested successfully considering different new transients. Overall, through prompt prediction of NPP behavior under different transients, the study aims at demonstrating the potential of artificial intelligence to empower rapid emergency response planning and risk mitigation strategies.

• Nuclear Engineering and Technology
• /
• v.14 no.2
• /
• pp.51-63
• /
• 1982

Stability of a PWR core against xenon-induced axial power oscillation is studied using one-dimensional xenon trausient analysis code, DD1D, that has been developed and verified at KAERI. Analyzed by DD1D utilizing the Kori Unit 1 design and operating data is the sensitivity of axial stability in a PWR core to the changes in core physical parameters including core power level, moderator temperature coefficient, core inlet temperature, doppler power coefficient and core average turnup. Through the sensitivity study the Kori Unit 1 core is found to be stable against axial xenon oscillation at the beginning of cycle 1. But, it becomes less stable as turnup progresses, and unstable at the end of the cycle. Such a decrease in stability is mainly due to combined effect of changes in axial power distribution, moderator temperature coefficient and doppler power coefficient as core turnup progresses. It is concluded from the stability analysis of the Kori Unit 1 core that design of a large PWR with high power density and increased dimension can not avoid xenon-induced axial power instabilities to some extents, especially at the end of cycle.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
• /
• v.1 no.4
• /
• pp.265-275
• /
• 1989

An experimental study of decaying swirl air flow has been obtained by tangential inlet in a straight tube with Reynolds number range 20,000~40,000. The friction factor, swirl angle, velocity profiles and turbulent intensity are measured by using micro-manometer and hot-wire anemometer. It is found that the swirl flow behaviors depend on the swirl intensity along the test tube.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
• /
• v.11 no.1
• /
• pp.81-90
• /
• 1999

The performance evaluation technique of a heat exchanger is described by using a transient response analysis for the determination of an average heat transfer coefficient. The model using a finite difference method can accommodate arbitrary inlet fluid temperature as well as longitudinal conduction. Temperature histories are obtained from the experiments at the inlet and outlet of test core. Heat transfer coefficient and friction factor of the plate array are obtained in short times using the data reduction program of transient response analysis in the single-blow method. The results agree very well with theoretical results. It is shown that the rms deviations are very small and the performance evaluation technique gives rapid and accurate results.