• Nuclear Engineering and Technology
• /
• 제44권7호
• /
• pp.709-718
• /
• 2012

The pressurized water reactor APR1400 adopts DVI (Direct Vessel Injection) for the emergency cooling water in the upper downcomer annulus. The International Standard Problem number 50 (ISP-50) was launched with the aim to investigate thermal hydraulic phenomena during a 50% DVI line break scenario with best estimate codes making use of the experimental data available from the ATLAS facility located at KAERI. The present work describes the calculation results obtained for the ISP-50 using the RELAP5/MOD3.3 system code. The work aims at validation and assessment of the code to reproduce the observed phenomena and investigate about its limitations to predict complicated mixing phenomena between the subcooled emergency cooling water and the two-phase flow in the downcomer. The obtained results show that the overall trends of the main test variables are well reproduced by the calculations. In particular, the pressure in the primary system show excellent agreement with the experiment. The loop seal clearance phenomenon was observed in the calculation and it was found to have an important influence on the transient progression. Moreover, the collapsed water levels in the core are accurately reproduced in the simulations. However, the drop in the downcomer level before the activation of the DVI from safety injection tanks was underestimated due to multi-dimensional phenomena in the downcomer that are not properly captured by one-dimensional simulations.

• 한국태양에너지학회 논문집
• /
• 제31권1호
• /
• pp.15-22
• /
• 2011

It is a great interest to convert more energy in the heat source into the power and to improve the efficiency of power generating processes. Since the efficiency of power generating processes becomes poorer as the temperature of the source decreases, to use an ammonia-water mixture instead of water as working fluid is a possible way to improve the efficiency of the system. In this work performance of ammonia-water regenerative Rankine cycle is investigated for the purpose of extracting maximum power from low-temperature waste heat in the form of sensible energy. Special attention is paid to the effect of system parameters such as mass fraction of ammonia and turbine inlet pressure on the characteristics of system. Results show that the power output increases with the mass fraction of ammonia in the mixture, however workable range of the mass fraction becomes narrower as turbine inlet pressure increases and is able to reach 16.5kW per unit mass flow rate of source air at $180^{\circ}C$.

• Nuclear Engineering and Technology
• /
• 제44권7호
• /
• pp.719-726
• /
• 2012

The object of this work is the validation and assessment of the TRACE v5.0 code using the scaled test ATLAS1 facility in the context of a DVI2 line break. In particular, the experiment selected models the 50%, 6-inch break of a DVI line. The same experiment was also adopted as a reference test in the ISP-503. The ISP-50 was proposed to, and accepted by, the OECD/NEA/CSNI due to its technical importance in the development of a best-estimate of safety analysis methodology for DVI line break accidents. In particular, the behavior of the two-phase flow in the upper annulus downcomer was expected to be complicated. What resulted was the need for relevant models to be implemented into safety analysis codes, in order to predict these thermal hydraulic phenomena correctly.

• Nuclear Engineering and Technology
• /
• 제53권5호
• /
• pp.1556-1568
• /
• 2021

The plate-type fuel assembly adopted in nuclear research reactor suffers from complicated effect induced by non-uniform irradiation, which might affect stress conditions, mechanical behaviors and thermal-hydraulic performance of the fuel assembly. This paper is the Part II work of a two-part study devoted to analyzing the complex unique mechanical deformation and thermal-hydraulic characteristics for the typical plate-type fuel assembly under irradiation effect, which is on the basis of developed and verified numerical thermal-fluid-structure coupling methodology under irradiation in Part I of this work. The mechanical deformation, thermal-hydraulic performance and Mises stress have been analyzed for the typical plate-type fuel assembly consisting of support plates under non-uniform irradiation. It was interesting to observe that: the plate-type fuel assembly including the fuel plates and support plates tended to bend towards the location with maximum fission rate; the hot spots in the fuel foil appeared at the location with maximum thickness increment; the maximum Mises stress of fuel foil was located at the adjacent location with the maximum plate thickness increment et al.

• Nuclear Engineering and Technology
• /
• 제49권4호
• /
• pp.675-691
• /
• 2017

Jets in cross flow are of fundamental industrial importance and play an important role in validating turbulence models. Two jet configurations related to thermal fatigue phenomena are investigated: ${\bullet}$ T-junction of circular tubes where a heated jet discharges into a cold main flow and ${\bullet}$ Rectangular jet marked by a scalar discharging into a main flow in a rectangular channel. The T-junction configuration is a classical test case for thermal fatigue phenomena. The Vattenfall T-junction experiment was already subject of an OECD/NEA benchmark. A LES modelling and calculation strategy is developed and validated on this data. The rectangular-jet configuration is important for basic physical understanding and modelling and has been analyzed experimentally at CEA. The experimental work was focused on turbulent mixing between a slightly heated rectangular jet which is injected perpendicularly into the cold main flow of a rectangular channel. These experiments are analyzed for the first time with LES. The overall results show a good agreement between the experimental data and the CFD calculation. Mean values of velocity and temperature are well captured by both RANS calculation and LES. The range of critical frequencies and their amplitudes, however, are only captured by LES.

• Nuclear Engineering and Technology
• /
• 제55권10호
• /
• pp.3754-3767
• /
• 2023

Turbulent jet discharges in subcooled water pools are essential for safety systems in nuclear power plants, specifically in the pressure suppression pool of boiling water reactors and In-containment Refueling Water Storage Tank of advanced pressurized water reactors. The gas and liquid flow in these systems is investigated using multiphase flow analysis. This field has been extensively examined using a combination of experiments, theoretical models, and Computational Fluid Dynamics (CFD) simulations. ANSYS CFX offers two approaches to model multiphase flow behavior. The non-homogeneous Eulerian-Eulerian Model has been used in this work; it computes global information and is more convenient to study interpenetrated fluids. This study utilized the Large Eddy Simulation Model as the turbulence model, as it is better suited for non-stationary and buoyant flows. The CFD results of this study were validated with experimental data and theoretical results previously obtained. The figures of merit dimensionless penetration length and the dimensionless buoyancy length show good agreement with the experimental measurements. Correlations for these variables were obtained as a function of dimensionless numbers to give generality using only initial boundary conditions. CFD numerical model developed in this research has the capability to simulate the behavior of non-condensable gases discharged in water.

• Nuclear Engineering and Technology
• /
• 제52권2호
• /
• pp.296-307
• /
• 2020

Permanent magnet sodium flowmeter (PMSF) have been used to measure the sodium flow in fast breeder reactors. Due to the effects of irradiation, thermal cycling, time lapse, etc., the magnetic flux density of the PMSF will decrease after being used in the reactor for a period of time. Therefore, it must be calibrated regularly. But some flowmeters that immersed in sodium cannot be removed for an off-line calibration, so the on-line calibration is required. However, the best online calibration accuracy of PMSF using cross-correlation analysis method was 2.0-level without considering the repeatability. In order to further improve this work, the operational principle of the transducer in PMSF is analyzed and the design principle of the transducer is proposed. The transducers were tested on the sodium flow loop to collect the experimental data. The signal characteristics are analyzed from the time and frequency domains, respectively. The cross-correlation analysis method based on biased estimation is adopted to obtain the flow rate. The verification experimental results showed that the measurement accuracy is 1.0-level when the flow velocity is above 0.5 m/s, and the measurement accuracy is 3.0-level when the flow velocity is in the range of 0.2 m/s to 0.5 m/s.

• Nuclear Engineering and Technology
• /
• 제39권4호
• /
• pp.299-312
• /
• 2007

The program on thermal-hydraulic evaluation by testing and analysis (THETA) for the development and licensing of the new design features in the APR1400 (Advanced Power Reactor-1400) is briefly introduced with a presentation on the research motivation and typical results of the separate effect tests and analyses of the major design features. The first part deals with multi-dimensional phenomena related to the safety analysis of the APR1400. One research area is related to the multidimensional behavior of the safety injection (SI) water in a reactor pressure vessel downcomer that uses a direct vessel injection type of SI system. The other area is associated with the condensation of steam jets and the resultant thermal mixing in a water pool; these phenomena are relevant to the depressurization of a reactor coolant system (RCS). The second part describes our efforts to develop new components for safety enhancements, such as a fluidic device as a passive SI flow controller and a sparger to depressurize the RCS. This work contributes to an understanding of the new thermal-hydraulic phenomena that are relevant to advanced reactor system designs; it also improves the prediction capabilities of analysis tools for multi-dimensional flow behavior, especially in complicated geometries.

• Nuclear Engineering and Technology
• /
• 제53권3호
• /
• pp.1020-1028
• /
• 2021

The nuclear reactor coolant pump transferring heat energy inherently brings with it the unsteady flow and inevitably threatens to the safe operation of the pump unit, especially with the pressure pulsation induced by the rotor-stator interaction. In this paper, the characteristics of pressure pulsation of the diffuser in a nuclear reactor coolant pump were investigated by the numerical simulation with experimental validation. Pressure pulsation signals measured synchronously from sensors mounted on the radial diffuser of a model pump were analyzed via Welch's method. Frequency components induced by the rotor-stator interaction can be revealed by the diameter mode analysis method. The pressure pulsation of the diffuser is dominated by the blade passing frequency and its harmonics, which are free from the effect of flow rate and rotational speed while the corresponding amplitudes are easily affected by different operational conditions and measuring positions. The non-uniformity is much more affected by the rotational speed than the flow rate. This research is helpful for further work to reduce the pressure pulsation for the reactor coolant pump.

• Nuclear Engineering and Technology
• /
• 제52권9호
• /
• pp.1998-2008
• /
• 2020

Solenoid operated valves (SOV) play important roles in industrial process to control the flow of fluids. Solenoid valves can be found in so many industries as well as the nuclear plant. The ability to be able to detect the presence of faults and predicting the remaining useful life (RUL) of the SOV is important in maintenance planning and also prevent unexpected interruptions in the flow of process fluids. This paper proposes a fault diagnosis method for the alternating current (AC) powered SOV. Previous research work have been focused on direct current (DC) powered SOV where the current waveform or vibrations are monitored. There are many features hidden in the AC waveform that require further signal analysis. The analysis of the AC powered SOV waveform was done in the time and frequency domain. A total of sixteen features were obtained and these were used to classify the different operating modes of the SOV by applying a machine learning technique for classification. Also, a deep neural network (DNN) was developed for the prediction of RUL based on the failure modes of the SOV. The results of this paper can be used to improve on the condition based monitoring of the SOV.