• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.38 no.1
• /
• pp.81-88
• /
• 2014

An experimental study is performed to obtain an average heat transfer coefficient around the perimeter of a near horizontal tube. For the test a stainless steel tube of 50.8 mm diameter submerged in water at atmospheric pressure is used. Both subcooled and saturated pool boiling conditions are considered and the inclination angle of the tube is changed from the horizontal position to $9^{\circ}$ in steps of $3^{\circ}$. In saturated water, the local boiling heat transfer coefficient at the azimuthal angle of $90^{\circ}$ from the tube bottom can be regarded as the average of the coefficients regardless of the tube inclination angles. However, when the water is subcooled the location for the average heat transfer coefficient depends on the inclination angle and the heat flux. It is explained that the major mechanisms changing the heat transfer are closely related with the intensity of the liquid agitation and the generation of big size bubbles through bubble coalescence.

• Journal of the Korean Society for Nondestructive Testing
• /
• v.16 no.1
• /
• pp.29-33
• /
• 1996

The eddy current test(ECT) probe of differential encircling coil type was designed and fabricated, and the optimum condition of ECT was derived for the examination of the irradiated fuel rod. The correlation between ECT test frequency and phase & amplitude was derived by performing the test of the standard rig that includes inner notches, outer notches and through-holes. The defect of through-hole was predicted by ECT at the G33-N2 fuel rod irradiated in the Kori-1 nuclear power reactor. The metallographic examination on the G33-N2 fuel rod was Performed at the defect location predicted by ECT. The result of metallographic examination for the G33-N2 fuel rod was in good agreement with that of ECT. This proves that the evaluation for integrity of irradiated fuel rod by ECT is reliable.

• Journal of Powder Materials
• /
• v.23 no.5
• /
• pp.384-390
• /
• 2016

Tri-isotropic (TRISO) coatings on zirconia surrogate beads are deposited using a fluidized-bed vapor deposition (FB-CVD) method. The silicon carbide layer is particularly important among the coated layers because it acts as a miniature pressure vessel and a diffusion barrier to gaseous and metallic fission products in the TRISO-coated particles. In this study, we obtain a nearly stoichiometric composition in the SiC layer coated at $1400^{\circ}C$, $1500^{\circ}C$, and $1400^{\circ}C$ with 20 vol.% methyltrichlorosilane (MTS), However, the composition of the SiC layer coated at $1300-1350^{\circ}C$ shows a difference from the stoichiometric ratio (1:1). The density decreases remarkably with decreasing SiC deposition temperature because of the nanosized pores. The high density of the SiC layer (${\geq}3.19g/cm^2$) easily obtained at $1500^{\circ}C$ and $1400^{\circ}C$ with 20 vol.% MTS did not change at an annealing temperature of $1900^{\circ}C$, simulating the reactor operating temperature. The evaluation of the mechanical properties is limited because of the inaccurate values of hardness and Young's modulus measured by the nano-indentation method.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.9 no.1
• /
• pp.15-19
• /
• 2013

The auxiliary feedwater is an important to remove the heat from the reactor core when the main feedwater system is unavailable. In most initiating events in Probabilistic Safety Assessment(PSA), the operaton of this system is required to mitigate the accidents. For one of domestic nuclear power plants, a design change of a turbine-driven auxiliary feedwater pump(TD-AFWP), pipe, and valves in the auxiliary system is implemented due to the aging related deterioration by long time operation. This change includes the replacement of the TD-AFWP, the relocation of some valves for improving the system availability, a new cross-tie line, and the installation of manual valves for maintenance. The design modification affects the PSA because the system is critical to mitigate the accidents. In this paper, the safety effect of the change of the auxiliary feedwater system is assessed with regard to the PSA view point. The results demonstrate that this change can supply the auxiliary feedwater from the TD-AFWP in the accident with the motor-driven auxiliary feedwater pump(MD-AFWP) unavailable due to test or maintenance. In addition, the change of MOV's normal position from "close" to "open" can deliver the water to steam generator in the loss of offsite power(LOOP) event. Therefore, it is confirmed that the design change of the auxiliary feedwater system reduces the total core damage frequency(CDF).

• Journal of Energy Engineering
• /
• v.22 no.2
• /
• pp.148-155
• /
• 2013

The steam generators of OPR-1000 plants have Alloy 600 and Alloy 690 as the tube material and its tube expansion method is the explosive expansion method. According to the experience of these plants, circumferential cracks were largely occurred in steam generator tubes expanded by the explosive expansion method and their locations were the outer surface of tube expansion transition region surrounding with piled-up sludge. But even though tubes have the same conditions, tubes with the hydraulic expansion method shows the prevail trend of axial cracks compared to circumferential cracks. Therefore in this study, in order to identify the difference of such phenomena as above, configurations of tube and tubesheet were modeled and at operating conditions, stress values applied in the tube expansion transition area in accordance with tube expansion methods were calculated by using computational program and the direction and the predominance of cracks were evaluated.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.5 no.1
• /
• pp.1-8
• /
• 2009

Nuclear hydrogen production using high temperature heat of a very high temperature reactor(VHTR) is one of the most attractive ways of mass hydrogen production without greenhouse gas emission. In many countries, sulfur-iodine(S-I) thermochemical process and high temperature steam electrolysis(HTSE) process are being investigated. In such processes, corrosion behavior of Intermediate heat exchanger materials are the most critical issues. Especially in a HTSE system, several heat exchangers will be facing hot steam conditions. In this paper, the status of high temperature corrosion researches in hot steam and supercritical water conditions are reviewed in view of the implication to HTSE conditions. Based on the review, test condition and plan of the hot steam corrosion of the candidate materials are formulated and described in some details along with the schematics of the test set-up. The test results and subsequent evaluation will be used in development of a interface system between the HTSE hydrogen production system and the VHTR.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.4 no.3
• /
• pp.209-216
• /
• 2006

Characteristics of heat release process, while the Wigner energy was drawn off the graphite during DSC(Differential Scanning Calorimenter) measurement as an example of annealing process which is one of release methods of Wigner energy that is contained in the irradiated graphite, was studied. Linear temperature rise method in DSC operation was selected to estimate the total Wigner energy content and the heat release rate of each graphite samples, which were located in several positions in the thermal column in KRR-2 research reactor. As an annealing process in DSC operation Wigner energy of the irradiated graphite samples were totally released by heat supplying to the graphite from room temperature to $500^{\circ}C$, in DSC. Characteristics of Wigner energy release from the graphite sample was well correlated with the various activation energy model of the kinetic equation.

• Korean Journal of Materials Research
• /
• v.14 no.10
• /
• pp.713-718
• /
• 2004

To study the dynamic strain aging behavior of Zr-0.4Sn-1.5Nb-0.2Fe sample tube for nuclear fuel cladding in the range of pressurized water reactor (PWR) operation temperature, the tensile tests of the tube specimens, which had been finally heat-treated at $470^{\circ}C\;and\;510^{\circ}C$, had been carried out with the strain rate $1.67{\times}10^{-2}/s\;and\;8.33{\times}10^{-5}/s$ at the various temperatures from room temperature to $500^{\circ}C$. It was observed that the elongation of the specimens got shortened as the temperature increased from $200^{\circ}C\;to\;340^{\circ}C$. The specimens that were finally heat-treated at $470^{\circ}C$ showed a plateau more remarkably on the plot of yield strength-temperature than those heat-treated at $510^{\circ}C$. In the range of $310\sim400^{\circ}C$, the strain rate sensitivity of the specimens finally heat-treated at $510^{\circ}C$ was $30.4\%\sim33.7\%$ lower but the work hardening exponent index of the specimens was a little higher than that without dynamic strain aging effect.

• Transactions of the Korean hydrogen and new energy society
• /
• v.26 no.2
• /
• pp.120-126
• /
• 2015

Recently, the ever-increasing use of fossil fuels for rapid industrial development and population significantly caused an environment pollution and global warming such as climate change. So research and development of sustainable and eco-friendly energy have been performed. Especially the interest in nuclear fusion fuel was significantly increased from the developed countries. The system of fusion energy production was tritium separation, storage and delivery, and purification. Republic of Korea is in charge of Storage and Delivery System (SDS) in the International Thermonuclear Experimental Reactor (ITER). Welding part of the SDS bottles for storing the tritium is known to be susceptible to hydrogen embrittlement. In this study, conducted a study for the relaxation of the stability and hydrogen embrittlement of the weld area. The hydrogen heat treatment was processed through the Pressure-Composition-Temperature (PCT) device according to the time variation. Also mechanical properties such as impact test and hardness test according to using the alkaline cleaning liquid for hydrogen embrittlement relief and the fracture was observed by scanning electron microscopy (SEM) after the mechanical properties evaluation.

• Transactions of the Korean hydrogen and new energy society
• /
• v.26 no.2
• /
• pp.114-119
• /
• 2015

Nowdays, fossil fuels have been used as an important resource in development of industry. But it is limited and caused climate change such as pollution and global warming. So nuclear fusion research is being issued with tritium to develop eco-friendly and sustainable energy. Republic of Korea is in charge of Storage and Delivery System (SDS) in the International Thermonuclear Experimental Reactor (ITER), weld present in the SDS bottles are easily exposed to the hydrogen embrittlement of special characteristics of the hydrogen in hydrogen atmosphere, When the hydrogen embrittlement is rapidly progresses, the cracking is generated in the weld zone. Due to this cracking, the risk of leakage of tritium into the atmosphere occurs. In this study, hydrogen heat treatment was processed through the Pressure-Composition-Temperature (PCT) device according to the time variation. Also mechanical properties such as rupture strength test, three point bend test and hardness test in accordance with the respective time have been conducted and the fracture was observed by scanning electron microscopy(SEM) after the mechanical properties evaluation.