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

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.108-113
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• 1996

Several analytical models for the departure from nucleate boiling (DNB) phenomenon have been developed during the last decade. Among these, Chang & Lee's model based on a bubble crowding mechanism is remarkable in the fundamental features characterized as the formulation of mass, energy, and momentum balance equation at thermal-hydraulic conditions leading to the DNB. However, Bricard and Souyri remarked that the assumption of stagnant bubbly layer at the DNB condition is questionable and the signs on the axial projections of the momentum fluxes at the core/bubbly layer interface in the momentum balance equations are erroneous. From this remark, Chang & Lee's model has been re-examined and modified by correcting the erroneous treatments in the momentum balance equations and removing the spurious assumptions. The revised model predicts well the extensive DNB data of water in uniformly heated tubes at low qualities and shows more accurate prediction compared with the original model.

• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.274-284
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• 1992

A 1/15-scale CREARE experiment, which simulates the thermal-hydraulic behavior in the reactor pressure vessel of a PWR during a hypothetical Loss Of Coolant Accident, has been analyzed using CATHARE code for the associated model assessment to represent the phenomenon. The key parameters examined in the CREARE experiment were known as ECC water injection rate. ECC water subcooling, system pressure, and steam flow rate coming out from the core bottom. The present CATHARE simulation, however, has been mainly focused on qualitative analysis of a countercurrent flow in the downcomer. The discrepancy of the simulation results with the experimental data is considered arising primarily from an inadequate numerical representation as well as an interfacial friction model. Accordingly it is suggested from the sensitivity studies that either multidimensional approach or further examination of momentum equations at a junction near a volume element in CATHARE be necessary in order to represent the phenomenon more realistically.

• 유체기계공업학회:학술대회논문집
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• 2006.08a
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• pp.489-492
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• 2006

The present work was performed to investigate the thermal and hydraulic characteristics of flow inside the plain and turbulator flat tubes for the automobile application. The pressure drop and heat transfer coefficient at laminar, transition and turbulent regimes were studied experimentally and numerically. The flow transition was confirmed by flow visualization and quantitative data. It is proposed equations for the friction and heat transfer coefficient in the fully developed laminar flow inside rectangular tube as function of aspect ratio.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.21 no.12
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• pp.695-702
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• 2009

Dynamic behavior of Libr-water absorption system using low-temperature hot water was investigated numerically. Thermal-hydraulic model of single-effect/double-lift 100 RT chiller was developed by applying transient conservation equations of total mass, Libr mass, energy and momentum to each component. Transient variations of system properties and transport variables were analysed during start-up operation. Numerical analysis were performed to quantify the effects of bulk concentration and part-load operation on the system performance in terms of cooling capacity, coefficient of performance, and time constant of system. For an absorption chiller considered in the present study, optimum bulk concentration was found to exist, which resulted in the minimum time constant with stable cooling capacity. COP and time constant increased as the load decreased down to 40%, below which the time constant increased abruptly and COP decreased as the load decreased further.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.16 no.10
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• pp.990-998
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• 2004

Dynamic behavior of an ammonia-water absorption system was investigated numerically. Thermal-hydraulic model for a single-effect 3 RT chiller was developed by applying transient conservation equations of total mass, $NH_3$ mass, energy and momentum to each component. Transient variations of system properties and transport variables were analysed during start-up operation. Numerical analyses were performed to quantify the effects of bulk concentration and charging ratio on the system performance in terms of cooling capacity, coefficient of performance, and time constant of system. For an absorption chiller considered in the present study, optimum charging ratio and bulk concentration were to found to exist, which resulted in the maximum cooling capacity and COP. The time constant increased as the charging ratio increased, but decreased with the increase of bulk concentration.

• Nuclear Engineering and Technology
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• v.42 no.6
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• pp.636-655
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• 2010

A thermal-hydraulic code, named CUPID, has been being developed for the realistic analysis of transient two-phase flows in nuclear reactor components. The CUPID code development was motivated from very practical needs, including the analyses of a downcomer boiling, a two-phase flow mixing in a pool, and a two-phase flow in a direct vessel injection system. The CUPID code adopts a two-fluid, three-field model for two-phase flows, and the governing equations are solved over unstructured grids with a semi-implicit two-step method. This paper presents an overview of the CUPID code development and assessment strategy. It also presents the code couplings with a system code, MARS, and, a three-dimensional reactor kinetics code, MASTER.

• Nuclear Engineering and Technology
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• v.35 no.5
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• pp.399-411
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• 2003

A model predictive control method is applied to design an automatic controller for thermal power control in a reactor core. The basic concept of the model predictive control is to solve an optimization problem for a finite future at current time and to implement as the current control input only the first optimal control input among the solutions of the finite time steps. At the next time step, the second optimal control input is not implemented and the procedure to solve the optimization problem is then repeated. The objectives of the proposed model predictive controller are to minimize the difference between the output and the desired output and the variation of the control rod position. The nonlinear PWR plant model (a nonlinear point kinetics equation with six delayed neutron groups and the lumped thermal-hydraulic balance equations) is used to verify the proposed controller of reactor power. And a controller design model used for designing the model predictive controller is obtained by applying a parameter estimation algorithm at an initial stage. From results of numerical simulation to check the controllability of the proposed controller at the $5\%/min$ ramp increase or decrease of a desired load and its $10\%$ step increase or decrease which are design requirements, the performances of this controller are proved to be excellent.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.24 no.8
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• pp.1046-1055
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• 2000

The flow and thermal phenomena in flat tubes of radiator are analyzed numerically. To predict the characteristics of heat transfer and pressure drop, the flow analysis program for three-dimensional complex geometry is developed, which adopted an non-staggered grid system and Cartesian velocities as dependent variables of the momentum equations. Using the developed program, the effect of tube specifications on the heat transfer characteristics is investigated for various flat tubes. From this study, the following results are obtained; (1) For the same hydraulic diameter($D_h{\doteq}5.2$mm), the Nusselt numbers of three basic modeis(D, J, and H-model) are 8.71, 8.92, and 10.58, respectively, and the pressure drops of D-, J-, and H-model are predicted as $-3.08{\times}10^{-2}\;Pa,\;-3.12{\times}10^{-2}\;Pa,\;and\; -3.98{\times}10^{-2}$ Pa, (2) In case of the same flat tube specification, the fins must be brazed at upper tube surface because the heat is more vividly transferred. Therefore, it is found that the H- model is the most effective tube as a heat exchanger and these results are used as a fundamental data for the design of tube.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.455-465
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• 2021

The Molten Salt Reactor (MSR), one of the six advanced reactor types of the 4th generation nuclear energy systems, has many impressive features including economic advantages, inherent safety and nuclear non-proliferation. This paper introduces a system analysis code named TREND, which is developed and used for the steady and transient simulation of MSRs. The TREND code calculates the distributions of pressure, velocity and temperature of single-phase flows by solving the conservation equations of mass, momentum and energy, along with a fluid state equation. Heat structures coupled with the fluid dynamics model is sufficient to meet the demands of modeling MSR system-level thermal-hydraulics. The core power is based on the point reactor neutron kinetics model calculated by the typical Runge-Kutta method. An incremental PID controller is inserted to adjust the operation behaviors. The verification and validation of the TREND code have been carried out in two aspects: detailed code-to-code comparison with established thermal-hydraulic system codes such as RELAP5, and validation with the experimental data from MSRE and the CIET facility (the University of California, Berkeley's Compact Integral Effects Test facility).The results indicate that TREND can be used in analyzing the transient behaviors of MSRs and will be improved by validating with more experimental results with the support of SINAP.