• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.29 no.8
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• pp.408-418
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• 2017

Supercritical carbon dioxide ($sCO_2$) power system is emerging technology because of its high cycle efficiency and compactness. Meanwhile, PCHE (Printed Circuit Heat Exchanger) is gaining attention in $sCO_2$ power system technology because PCHE with high pressure-resistance and larger heat transfer surface per unit volume is fundamentally needed. Thermo-fluidic characteristics of $sCO_2$ in the micro channel of PCHE should be investigated. In this study, heat transfer characteristics of $sCO_2$ of various inlet conditions and cross-sectional shapes of single micro channel were investigated experimentally. Experiment was conducted at supercritical state of higher than critical temperature and pressure. Test sections were made of copper and hydraulic diameter was 1 mm. Convective heat transfer coefficients were measured according to each interval of the channel and pressure drop was also measured. Convective heat transfer coefficients from experimental data were compared with existing correlation. In this study, using measured data, a new empirical correlation to predict near critical region heat transfer coefficient is developed and suggested. Test results of single channel will be used for design of PCHE.

• Nuclear Engineering and Technology
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• v.40 no.4
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• pp.255-268
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• 2008

The greater demand for electricity and the available capacity within safety margins in some operating NPPs are prompting nuclear utilities to request license modification to enable operation at a higher power level, beyond their original license provisions. Such plant modifications require an in-depth safety analysis to evaluate the possible safety impact. The analysis must consider the thermo hydraulic, radiological and structural aspects, and the plant behavior, while taking into account the capability of the structures, systems and components, and the reactor protection and safeguard systems set points. The purpose of this paper is to introduce international experiences and approaches for implementation of power uprates related to the reactor thermal power of nuclear power plants. The paper is intended to give the reader a general overview of the major processes, work products, issues, challenges, events, and experiences in the power uprates program. The process of increasing the licensed power level of a nuclear power plants is called a power uprate. One way of increasing the thermal output from a reactor is to increase the amount of fissile material in use. It is also possible to increase the core power by increasing the performance of the high power bundles. Safety margins can be maintained by either using fuels with a higher performance, or through the use of improved methods of analysis to demonstrate that the required margins are retained even at the higher power levels. The paper will review all types of power uprates, from small to large, and across various reactor types, including light and heavy water, pressurized, and boiling water reactors. Generally, however, the content of the report focuses on power uprates of the stretch and extended type. The International Atomic Energy Agency (IAEA) is developing a technical guideline on power uprates and side effects of power uprates in nuclear power plants.