• The Transactions of the Korean Institute of Electrical Engineers
• /
• v.31 no.10
• /
• pp.109-115
• /
• 1982

Reactor dynamics were studied by reactor noise at TRIGA MARK Il reactor whose rated power is 250KW thermal. The power spectral densities(PSD) of the noise were measured by stochastic method with high resolution digital filters and Fast Fourier Transformers. The transfer function of the reactor at zero power was identical to the theoretical characteristics. When the power was increasec above 1KW, reactor showed its poswer resonances at 3Hz and 10 Hz. It was analyzed that 3Hz peak was generated by heat transfer and coolant flow effects and 10Hz peak by nuclear reaction effects.

• Nuclear Engineering and Technology
• /
• v.53 no.6
• /
• pp.1731-1735
• /
• 2021

The operation of the nuclear reactor requires accurate and fast methods and techniques for analysing its kinetics. These techniques become even more important when the MOX-fuel is used due to the lower value of delayed neutron fraction 𝛽 for ²³⁹Pu. Based on a Birth-and-Death process review, the mathematical model of thermal reactor core has been proposed different from existing ones. The analytical method for thermal point-reactor parameters evaluation is described within this work. The proposed method is applied for analysis of the unsteady transient processes taking place in a thermal reactor at its start-up or shutdown power change, as well as during small accidental power variation from the rated value. Theoretical determination of MASURCA reactor core reactivity through the analysis of experimental data on neutron time spectra was made.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
• /
• 2001.11a
• /
• pp.251-256
• /
• 2001

• Nuclear Engineering and Technology
• /
• v.51 no.4
• /
• pp.1017-1023
• /
• 2019

The increase of the alternate current frequency results in increased rotational speed of the electrical motors and connected pumps. The consequence for the reactor coolant pumps is increased flow in primary coolant system. Increase of the current frequency can be initiated by the subsynchronous resonance phenomenon (SSR). This paper analyses the implications of the SSR and consequential increase of the frequency on the nuclear power plant safety. The Simulink $MATLAB^{(R)}$ model of the steam turbine and governor system and RELAP5 computer code of the pressurized water reactor are used in the analysis. The SSR results in fast increase of reactor coolant pumps speed and flow in the primary coolant system. The turbine trip value is reached in short time following SSR. The increase of flow of reactor coolant pumps results in increase of heat removal from reactor core. This results in positive reactivity insertion with reactor power increase of 0.5% before reactor trip is initiated by the turbine trip. The main parameters of the plant did not exceed the values of reactor trip set points. The pressure drop over reactor core is small discarding the possibility of core barrel lift.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.21 no.4
• /
• pp.571-576
• /
• 2023

The initial development plans for the six reactor designs, soon after the release of Generation IV International Forum (GIF) TRM in 2002, were characterized by high ambition [1]. Specifically, the sodium-cooled fast reactor (SFR) and very-high temperature reactor (VHTR) gained significant attention and were expected to reach the validation stage by the 2020s, with commercial viability projected for the 2030s. However, these projections have been unrealized because of various factors. The development of reactor designs by the GIF was supposed to be influenced by events such as the 2008 global financial crisis, 2011 Fukushima accident [2, 3], discovery of extensive shale oil reserves in the United States, and overly ambitious technological targets. Consequently, the momentum for VHTR development reduced significantly. In this context, the aims of this study were to compare and analyze the development progress of the six Gen IV reactor designs over the past 20 years, based on the GIF roadmaps published in 2002 and 2014. The primary focus was to examine the prospects for the reactor designs in relation to spent nuclear fuel burning in conjunction with small modular reactor (SMR), including molten salt reactor (MSR), which is expected to have spent nuclear fuel management potential.

• Nuclear Engineering and Technology
• /
• v.53 no.8
• /
• pp.2753-2760
• /
• 2021

This paper proposes using sodium-cooled fast reactor technologies for use in hydrogen vapor methane (SMR) modification. Using three independent energy rings in the Russian BN-600 fast reactor, steam is generated in one of the steam-generating cycles with a pressure of 13.1 MPa and a temperature of 505 ℃. The reactor's second energy cycles can increase the gas-steam mixture's temperature to the required amount for efficient correction. The 620 ton/hr 540 ℃ steam generated in this cycle is sufficient to supply a high-temperature synthesis current source (700 ℃), which raises the steam-gas mixture's temperature in the reactor. The proposed technology provides a high rate of hydrogen production (approximately 144.5 ton/hr of standard H₂), also up to 25% of the original natural gas, in line with existing SMR technology for preparing and heating steam and gas mixtures will be saved. Also, exergy analysis results show that the plant's efficiency reaches 78.5% using HTR heat for combined hydrogen and power generation.

• Nuclear Engineering and Technology
• /
• v.55 no.6
• /
• pp.2047-2052
• /
• 2023

Risk-informed approach has been widely applied in the safety design, regulation, and operation of nuclear reactors. It has been commonly accepted that risk-informed design optimization should be used in the innovative reactor designs to make nuclear system highly safe and reliable. In spite of the risk-informed approach has been used in some advanced nuclear reactors designs, such as Westinghouse IRIS, Gen-IV sodium fast reactors and lead-based fast reactors, the process of risk-informed design of nuclear reactors is hardly to carry out when passive system reliability should be integrated in the framework. A practical method for new passive safety reactors based on probabilistic safety assessment (PSA) and passive system reliability analyze linking is proposed in this paper. New three-dimension frequency-consequence curve based on risk concept with three variables is used in this method. The proposed method has been applied to the determination optimization of design options selection in a 10 MWth lead-based research reactor(LR) to obtain one optimized system design in conceptual design stage, using the integrated reliability and probabilistic safety assessment program RiskA, and the computation resources and time consumption in this process was demonstrated reasonable and acceptable.

• Proceedings of the Korean Society of Precision Engineering Conference
• /
• 2004.10a
• /
• pp.1327-1330
• /
• 2004

An underwater robotic system has been developed and applied to visual inspection of reactor vessel internals. The Korea Electric Power Robot for Visual Test (KeproVt) consists of an underwater robot, a vision processor-based measuring unit, a master control station and a servo control station. The robot guided by the control station with the measuring unit can be controlled to have any motion at any position in the reactor vessel with $\pm$1 cm positioning and $\pm$2 degrees heading accuracies with enough precision to inspect reactor internals. A simple and fast installation process is emphasized in the developed system. The developed robotic system was successfully deployed at the Younggwang Nuclear Unit 1 for the visual inspection of reactor internals.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.16 no.3
• /
• pp.359-367
• /
• 2018

Long-lived Small Modular Reactors are being promoted as an innovative way of catering to emerging markets and isolated regions. They can be operated continuously for decades without requiring additional fuel. A novel configuration of long-lived reactor core employs a mixed neutron spectrum, providing an improvement in nonproliferation metrics and in safety characteristics. Starting with a base sodium reactor design, moderating material is inserted in outer core assemblies to modify the fast spectrum. The assemblies are shuffled once during core lifetime to ensure that every fuel rod is exposed to the thermalized spectrum. The Mixed Spectrum Reactor is able to maintain a core lifetime over two decades while ensuring the plutonium it breeds is below the weapon-grade limit at the fuel discharge. The main drawbacks of the design are higher front-end fuel cycle costs and a 58% increase in core volume, although it is alleviated to some extent by a 48% higher power output.

• Journal of Korean Society of Water and Wastewater
• /
• v.21 no.1
• /
• pp.83-89
• /
• 2007

A biological anoxic-aerobic reactor filled with porous media was operated in lab scale for the advanced wastewater treatment. The experiments were conducted for 6 months with three HRTs (4, 6, 8hr) and temperature of $23{\sim}25^{\circ}C$. Some other experimental conditions were as follows; nitrification reactor (MLSS 4,500mg/L, DO 3.3mg/L, $23{\sim}28^{\circ}C$), denitrification reactor(MLSS 8,000mg/L, ORP -100mV, Temp.$19{\sim}23^{\circ}C$). Average removal efficiencies of SS, $BOD_5$, $COD_{Cr}$, T-N, and T-P were 97.8%, 95.5%, 94.5%, 80.2%, and 60.6%, respectively. The reactor filled with porosity media showed stable removal capacity for organics and nutrients. Fast and complete nitrification and denitrification were accomplished. Maintaining high MLSS with porous media in the nitrification and denitrification reactor appears to enhance the nitrogen removal process. For the higher T-P removal, some coagulant addition process will be needed.