• Journal of the Korea Academia-Industrial cooperation Society
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• v.18 no.4
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• pp.15-20
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• 2017

Currently, residential air conditioners operate as a heat pump during winter. In this case, the outdoor heat exchanger acts as an evaporator obtaining heat from cold air. On the other hand, it acts as a condenser during summer transferring heat to hot air. The outdoor temperature changes significantly from high to low. Generally, the air-side j and f factors are obtained at a standard outdoor temperature. Therefore, the applicability of the j and f factors under different outdoor conditions needs to be checked. In this study, tests were conducted for a two-row louver finned heat exchanger changing the outdoor temperature to subzero. The effects of the tube-side brine flow rate were also checked. The results showed that air-side j and f factors were essentially constant and independent of the outdoor temperature, suggesting that an extension of j and f factors obtained under standard conditions to a low outdoor temperature is acceptable. All j and f factors agreed within 9% and 3%, respectively. Tests were also conducted by changing the coolant flow rate. Both the j and f factors did not change according to the flow rate, suggesting that the tube-side heat transfer correlation is acceptable.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.1
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• pp.74-83
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• 2020

A number of so-called 'Type IV' cracking was reported to occur at the welded joints of the P91 steel or P92 steel reheat steam piping systems in Korean supercritical thermal power plants. The reheat steam piping systems are subjected to severe thermal and pressure loading conditions of coolant higher than 570℃ and 4MPa, respectively. In this study, piping analyses and design evaluations were conducted for the piping system of a specific thermal plant in Korea and suggestions were made how structural integrity could be improved so that type IV cracks at the welded joints could be prevented. Integrity evaluations were conducted as per ASME B31.1 code with implicit consideration of creep effects which was used in original design of the piping system and as per nuclear-grade RCC-MRx code with explicit consideration of creep effects. Comparisons were made between the evaluation results from the two design rules. Another approach with modification or reduction of the redundant supports in the piping systems was investigated as a tool to mitigate thermal stresses which should essentially contribute to prevention of Type IV cracking without major modification of the existing piping systems. In addition, a post weld heat treatment method and repair weld method which could improve integrity of the welded joint of P91 steel were investigated.

• Nuclear Engineering and Technology
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• v.43 no.3
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• pp.301-308
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• 2011

The diametral creep of pressure tubes (PTs) in CANDU (CANada Deuterium Uranium) reactors is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of the heat transport system (HTS). PT diametral creep leads to diametral expansion, which affects the thermal hydraulic characteristics of the coolant channels and the critical heat flux (CHF). The CHF is a major parameter determining the critical channel power (CCP), which is used in the trip setpoint calculations of regional overpower protection (ROP) systems. Therefore, it is essential to predict PT diametral creep in CANDU reactors. PT diametral creep is caused mainly by fast neutron irradiation, temperature and applied stress. The objective of this study was to develop a bundle position-wise linear model (BPLM) to predict PT diametral creep employing previously measured PT diameters and HTS operating conditions. The linear model was optimized using a genetic algorithm and was devised based on a bundle position because it is expected that each bundle position in a PT channel has inherent characteristics. The proposed BPLM for predicting PT diametral creep was confirmed using the operating data of the Wolsung nuclear power plant in Korea. The linear model was able to predict PT diametral creep accurately.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3775-3786
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• 2023

Mixing Vane Grid (MVG) is one of the most important structures in fuel assembly due to its high performance in mixing the coolant and ultimately increasing Critical Heat Flux (CHF), which avoids the temperature rising suddenly of fuel rods. To evaluate the mixing performance of the MVG, a Total Diffusion Coefficient (TDC) mixing coefficient is defined in the subchannel analysis code. Conventionally, the TDC of the spacer grid is obtained from the combination of experiments and subchannel analysis. However, the processing of obtaining and determine a reasonable TDC is much challenging, it is affected by boundary conditions and MVG geometries. In is difficult to perform all the large and costing rod bundle tests. In this paper, the CFD method was applied in TDC analysis. A typical 5 × 5 MVG was simulated and validated to estimate the mixing performance of the MVG. The subchannel code was used to calculate the TDC. Firstly, the CFD method was validated from the aspect of pressure drop and lateral temperature distribution in the subchannels. Then the effect of boundary conditions including the inlet temperature, inlet velocities, heat flux ratio between hot and cold rods and the arrangement of hot and cold rods on MVG mixing and TDC were studied. The geometric effects on mixing are also carried out in this paper. The effect of vane pattern on mixing was investigated to determine which one is the best to represent the grid's mixing performance.

• Journal of Biosystems Engineering
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• v.12 no.3
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• pp.7-16
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• 1987

This study was carried out to develop a microcomputer-based data acquisition and control system which was able to collect the data of engine performance test automatically and control the speed and load of the engine. The results of the study are summarized as follows: 1. The signal processing devices, which were able to measure cylinder pressure, coolant temperature, compositions of exhaust gas, fuel consumption, engine rpm and torque etc., were developed. The results of calibration showed that all of devices had high accuracy ranging from 0.3% to 0.69% respectively. 2. The PIA (peripheral interface adapter) for interfacing digital signal and PTM (programmable timer module) for displaying real time every 0.0408 sec were designed and developed. 3. An engine-speed control system using a stepping motor and driver was developed. The control system had the stability, and faster settling time than the manual control system. 4. The automatic control system of electrical dynamometer, which was able to control the speed and load of dynamometer, was developed with a SSD (shackleton system driver) and D/A converter. 5. The computer programs, which were able to collect and process the data of engine tests, were developed using both the machine language and BASIC.

• Nuclear Engineering and Technology
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• v.41 no.6
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• pp.753-764
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• 2009

JAEA started the LSTF experiments in 1985 for the fourth stage of the ROSA Program (ROSA-IV) for the LWR thermal-hydraulic safety research to identify and investigate the thermal-hydraulic phenomena and to confirm the effectiveness of ECCS during small-break LOCAs and operational transients. The LSTF experiments are underway for the ROSA-V Program and the OECD/NEA ROSA Project that intends to resolve issues in thermal-hydraulic analyses relevant to LWR safety. Six types of the LSTF experiments have been done for both the system integral and separate-effect experiments among international members from 14 countries. Results of four experiments for the ROSA Project are briefly presented with analysis by a best-estimate (BE) code and a computational fluid dynamics (CFD) code to illustrate the capability of the LSTF and codes to simulate the thermal-hydraulic phenomena that may appear during SBLOCAs and transients. The thermal-hydraulic phenomena dealt with are coolant mixing and temperature stratification, water hammer up to high system pressure, natural circulation under high core power condition, and non-condensable gas effect during asymmetric SG depressurization as an AM action.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2395-2406
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• 2023

To study the influence of parameter uncertainty in small pressurized water reactor (SPWR) once-through steam generator (OTSG), the nonlinear mathematical model of the SPWR is firstly established. Including the reactor core model, the OTSG model and the pressurizer model. Secondly, a control strategy that both the reactor core coolant average temperature and the secondary-side outlet pressure of the OTSG are constant is adopted. Then, the uncertainty quantification method is established based on Latin hypercube sampling and statistical method. On this basis, the quantitative platform for parameter uncertainty of the OTSG is developed. Finally, taking the uncertainty in primary-side flowrate of the OTSG as an example, the platform application work is carried out under the variable load in SPWR and step disturbance of secondary-side flowrate of the OTSG. The results show that the maximum uncertainty in the critical output parameters is acceptable for SPWR.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.832-845
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• 2024

Reactor cavity cooling systems (RCCSs) comprising passive safety features use the atmosphere as a coolant, which cannot be lost. However, their drawback is that they are easily affected by atmospheric disturbances. To realize the commercial application of the two types of passive RCCSs, namely RCCSs based on atmospheric radiation and atmospheric natural circulation, their safety must be evaluated, that is, they must be able to remove heat from the reactor pressure vessel (RPV) surface at all times and under any condition other than under normal operating conditions. These include both expected and unexpected natural phenomena and accidents. Moreover, they must be able to eliminate the heat leakage emitted from the RPV surface during normal operation. However, utilizing all of the heat emitted from the RPV surface increases the degree of waste heat utilization. This study aims to understand the characteristics and degree of passive safety features for heat removal by comparing RCCSs based on atmospheric radiation and atmospheric natural circulation under the same conditions. It was concluded that the proposed RCCS based on atmospheric radiation has an advantage in that the temperature of the RPV could be stably maintained against disturbances in the ambient air.

• Journal of Energy Engineering
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• v.12 no.2
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• pp.74-83
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• 2003

In the advanced nuclear power plants including APR1400, the SDVS (Safety Depressurization and Vent System) is adopted to increase the plant safety using the concept of feed-and-bleed operation. In the case of the TLOFW (Total Loss of Feedwater), the POSRV (Power Operated Safety Relief Value) located at the top of the pressurizer is expected to open due to the pressurization of the reactor coolant system and discharges steam and/or water mixture into the water pool, where the mixture is condensed. During the condensation of the mixture, thermal-hydraulic loads such as pressure and temperature variations are induced to the pool structure. For the pool structure design, such thermal-hydraulic aspects should be considered. Understanding the phenomena of the submerged steam jet condensation in a water pool is helpful for system designers to design proper pool structure, sparger, and supports etc. This paper reviews and evaluates the steam jet condensation in a water pool on the physical phenomena of the steam condensation including condensation regime map, heat transfer coefficient, steam plume, steam jet condensation load, and steam jet induced flow.

• Nuclear Engineering and Technology
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• v.12 no.3
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• pp.153-162
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• 1980

The investigation of the fuel cladding temperature behavior and heat transfer mechanism during the reflooding phase of a LOCA plays an important role in performance evaluation of ECCS and safety analysis of water reactors. Reflooding experiments were performed with horizontal and vertical flow channels to investigate the effect of coolant flow channel orientation on rewetting process. Emphasis was mainly placed on the CANDU reactor which has horizontal pressure tubes in core, and the results were compared with those of vertical channel. Also to investigate the rewetting process visually, the experiments by using a rod in annulus and a quartz tube heated outside were performed. It can be concluded that the rewetting velocity in horizontal flow channel is clearly affected by flow stratification, however, the average rewetting velocity is similar to those in vertical flow channel for same conditions.