• Nuclear Engineering and Technology
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• v.35 no.4
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• pp.261-273
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• 2003

The characteristics of the recirculation operation of LMR which are different from the conventional plants such as PWR and fossil fuel plants were investigated using a computer code TSGS developed in this study. For simulating the transient behavior of the steam generation system, a water level control algorithm utilizing digital control hardware features was introduced. By investigation, the function of the recirculation operation was defined, the major features of the operation were found. Also good performance of the level control algorithm was confirmed.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2000.06a
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• pp.1829-1835
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• 2000

The experimental results of sodium-water reaction noise measurement in frequency range $1{/sim}200kHz$ are presented. The experiments of noise generation under the condition of sodium test facility, water leak rate $0.01{\sim}1.2g/s$ and temperature of sodium $250{\sim}500^{\circ}C$, were carried out. From theoretical study it is noted that the noise resonant attenuation on hydrogen bubbles in liquid sodium plays the significant role for leak noise spectra formation. Interaction of leak noise and hydrogen bubbles in sodium being accompanied by thermal, emission and viscosity energy dissipation was studied. Acoustic noise spectra were investigated from point of view of water leak detection in sodium/water steam generator. The results of sodium-water reaction noise absorption on hydrogen bubbles in liquid sodium by temperature $250{\sim}500^{\circ}C$ are presented. The theoretical model of noise absorption using the coefficients of attenuation was developed. From calculation the coefficients of attenuation were estimated.

• Nuclear Engineering and Technology
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• v.47 no.1
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• pp.47-58
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• 2015

Four recycling scenarios involving pyroprocessing of spent fuel (SF) have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR), KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU) SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. Thefsensitivity of cooling time before prior to pyro-processing was studied. As the cooling time sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC) decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs). If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE) isotopes. The RE isotope recovery factor should be lowered to ${\leq}20%$ in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

• Nuclear Engineering and Technology
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• v.41 no.4
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• pp.427-446
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• 2009

In order to meet the increasing demand for electricity, Korea has to rely on nuclear energy due to its poor natural resources. In order for nuclear energy to be expanded in its utilization, issues with uranium supply and waste management issues have to be addressed. Fast reactor system is one of the most promising options for electricity generation with its efficient utilization of uranium resources and reduction of radioactive waste, thus contributing to sustainable development. The Korea Atomic Energy Research Institute (KAERI) has been performing R&Ds on Sodium-cooled Fast Reactors (SFRs) under the national nuclear R&D program. Based on the experiences gained from the development of KALIMER conceptual designs of a pool-type U-TRU-10%Zr metal fuel loaded reactor, KAERI is currently developing Advanced SFR design concepts that can better meet the Generation IV technology goals. This also includes developing, Advanced SFR technologies necessary for its commercialization and basic key technologies, aiming at the conceptual design of an Advanced SFR by 2011. KAERI is making R&D efforts to develop advanced design concepts including a passive decay heat removal system and a supercritical $CO_2$ Brayton cycle energy conversion system, as well as developing design methodologies, computational tools, and sodium technology. The long-term Advanced SFR development plan will be carried out toward the construction of an Advanced SFR demonstration plant by 2028.

• Journal of the Earthquake Engineering Society of Korea
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• v.2 no.4
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• pp.73-86
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• 1998

The purpose of this paper is to evaluate the shear stiffness, hysteretic behavior, and ultimate behavior of HDRB(High Damping Rubber Bearing), which will be included in the seismic isolation design guideline as requirements. To do this, two 1/8 scaled HDRB are designed, fabricated, and tested to show the mechanical characteristics. The shear stiffness obtained from the proposed equation of the shear stiffness shows a good agreement with those of the experiments. For analysis of the hysteretic behavior of HDRB using the modified rate model, the parameter equations are obtained from the experiments. Using the obtained parameter equations for the modified rate model, the seismic response analyses are carried out for 1-D system. The results of analysis well follow the hysteretic behavior of HDRB obtained from the experiments. To evaluate the ultimate behavior of HDRB used in this paper, the analyses are carried out using the modified macro model, which can consider the large shear deflection. The critical shear strain(CSS) is defined to express the maximum allowable shear strain and vertical load. From the analyses, the CSS, showing the instability, decreases significantly as increased the vertical loads. The CSS is not appeared for the design vertical load in the used HDRB. In analysis using about 5 times of design vertical load, the HDRB start to show the instability transient and for about 7 times, the CSS is about 350%.