• Nuclear Engineering and Technology
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• v.36 no.6
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• pp.528-539
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• 2004

It is essential in commercial reactors that the safety limits imposed on the fuel pellets and fuel clad barriers, such as the linear power density (LPD) and the departure from nucleate boiling ratio (DNBR), are not violated during reactor operations. In order to accurately monitor the safety limits of current reactor states, a detailed three-dimensional (3D) core power distribution should be estimated from the in-core detector signals. In this paper, we propose a calculation methodology for detailed 3D core power distribution, using in-core detector signals and core monitoring constants such as the 3D Coupling Coefficients (3DCC), node power fraction, and pin-to-node factors. Also, the calculation method for several core safety parameters is introduced. The core monitoring constants for the real core state are promptly provided by the core design code and on-line MASTER (Multi-purpose Analyzer for Static and Transient Effects of Reactors), coupled with the core monitoring program. through the plant computer, core state variables, which include reactor thermal power, control rod bank position, boron concentration, inlet moderator temperature, and flow rate, are supplied as input data for MASTER. MASTER performs the core calculation based on the neutron balance equation and generates several core monitoring constants corresponding to the real core state in addition to the expected core power distribution. The accuracy of the developed method is verified through a comparison with the current CECOR method. Because in all the verification calculation cases the proposed method shows a more conservative value than the best estimated value and a less conservative one than the current CECOR and COLSS methods, it is also confirmed that this method secures a greater operating margin through the simulation of the YGN-3 Cycle-1 core from the viewpoint of the power peaking factor for the LPD and the pseudo hot pin axial power distribution for the DNBR calculation.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1472-1482
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• 2017

Zirconium alloy cladding tube specimens were irradiated at $380^{\circ}C$ up to a fast neutron fluence of $7.5{\times}10^{24}n/m^2$ in a research reactor to investigate the effect of neutron irradiation on hydride reorientation and mechanical property degradation. Cool-down tests from $400^{\circ}C$ to $200^{\circ}C$ under 150 MPa tensile hoop stress were performed. These tests indicate that the irradiated specimens generated a smaller radial hydride fraction than did the unirradiated specimens and that higher hydrogen content generated a smaller radial hydride fraction. The irradiated specimens of 500 ppm-H showed smaller ultimate tensile strength and plastic strain than those characteristics of the 250 ppm-H specimens. This mechanical property degradation caused by neutron irradiation can be explained by tensile hoop stress-induced microcrack formation on the hydrides in the irradiation-damaged matrix and subsequent microcrack propagation along the hydrides and/or through the matrix.

• Proceedings of the KIEE Conference
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• 2009.07a
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• pp.932_933
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• 2009

A technology based on thyristors will be used to manufacture the super-conducting coil AC/DC converters because of the low ratio of cost over installed power compared to a design based on GTO or similar technology. But phase-controlled converter suffers from fundamental disadvantage. They inject current harmonics into the input ac mains due to their nonlinear characteristics, thereby distort the supply voltage waveform, and demand reactive power from the associated ac power system at retarded angles. To overcome this disadvantage, in the case of two series converters at the DC side, connected to the same step-down transformer, apply for the sequence control. It is the most simple and efficient way to reduce the reactive power consumption at low cost. Analytical sequence control algorithm is suggested, the validity of the proposed scheme has been verified by experimental results with the small-scaled International Thermonuclear Experimental Reactor (ITER) Power Supply to minimize reactive power consumption.

• Nuclear Engineering and Technology
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• v.27 no.3
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• pp.393-402
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• 1995

The structural analysis of the reactor coolant system mainly consist of too fields. The one is the static analysis considering the impact of pressure and temperature built up during normal operation. The other is the dynamic analysis to estimate the impact of postulated events such as the seismic loads or postulated branch line pipe breaks event. Since the most important goal of the RCS structural analysis is to prove the safety of the RCS during normal operation or postulated events, a widely proven theory having enough conservatism is adopted. The load occurring on the RCS during normal operation is considered as the basic design loading condition throughout whole plant life time. The most typical characteristic of the RCS during normal operation is the thermal expansion of the RCS caused by reactor coolant with high temperature and pressure. Therefore, the exact estimation on the thermal movement of the RCS is needed to get more clear understanding on the thermal movement behavior of the RCS. In this study, the general structural analysis concept and modeling method to evaluate the thermal movement of the RCS under the normal plant operation condition are presented. To discuss the validation of the suggested analysis, analysis results are compared with the measured data which ore referred from the standardized 1000 MWe PWR plant under construction.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.249-256
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• 2019

The 9Cr-1Mo ferritic-martensitic ODS steel is a promising structural material for the next generation nuclear power plants including fast reactors for application in reactor vessels and nuclear fuel. The ODS steel was cooled down by furnace cooling, air cooling, oil quenching and water quenching, respectively, after normalizing it at $1150^{\circ}C$ for 1 h and then tempering at $780^{\circ}C$ for 1 h. It is found that grain size, a relative portion of ferrite and martensite, martensitic lath configuration, behaviors of carbide precipitates, and hardness of the ODS steel are strongly dependent on a cooling rate. The grain size and martensitic lath width become smaller with the increase in a cooling rate. The carbides were precipitated at the grain boundaries formed between the ferrite and martensite phases and at the martensitic lath interfaces. In addition, the carbide precipitates become smaller and more widely dispersed with the increase in a cooling rate, resulting in that the faster cooling rate generated the higher hardness of the ODS steel.

• Journal of Sensor Science and Technology
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• v.18 no.5
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• pp.402-407
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• 2009

In this study, plastic optical fiber sensors which can measure thermal neutron dose in a mixed neutron-gamma field are developed and characterized. Using $^{252}Cf$ and $^{60}Co$ sources, the scintillators suitable for thermal neutron detection, are tested and the scintillating lights generated from a plastic optical fiber sensor in the Kyoto University Critical Assembly (kuca) core are measured. Also, the distributions of thermal neutron and gamma-ray are measured in a mixed field as a function of the distance from the center of the reactor core at KUCA and the distribution of thermal neutron is obtained using a subtraction method. Sensitivity of the fiber-optic radiation sensor system is about 0.49 V/mW according to power of the KUCA core and its relative error is about 1.2 %.

• Nuclear Engineering and Technology
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• v.41 no.4
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• pp.493-520
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• 2009

During the last four decades, 16 Pressurized Water Reactors (PWR) and 4 Pressurized Heavy Water Reactors (PHWR) have been constructed and operated in Korea, and nuclear fuel technology has been developed to a self-reliant state. At first, the PWR fuel design and manufacturing technology was acquired through international cooperation with a foreign partner. Then, the PWR fuel R&D by Korea Atomic Energy Research Institute (KAERI) has improved fuel technology to a self-reliant state in terms of fuel elements, which includes a new cladding material, a large-grained $UO_2$ pellet, a high performance spacer grid, a fuel rod performance code, and fuel assembly test facility. The MOX fuel performance analysis code was developed and validated using the in-reactor test data. MOX fuel test rods were fabricated and their irradiation test was completed by an international program. At the same time, the PWR fuel development by Korea Nuclear Fuel (KNF) has produced new fuel assemblies such as PLUS7 and ACE7. During this process, the design and test technology of fuel assemblies was developed to a self-reliant state. The PHWR fuel manufacturing technology was developed and manufacturing facility was set up by KAERI, independently from the foreign technology. Then, the advanced PHWR fuel, CANFLEX(CANDU Flexible Fuelling), was developed, and an irradiation test was completed in a PHWR. The development of the CANFLEX fuel included a new design of fuel rods and bundles.. The nuclear fuel technology in Korea has been steadily developed in many national R&D programs, and this advanced fuel technology is expected to contribute to a worldwide nuclear renaissance that can create solutions to global warming.

• KSBB Journal
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• v.20 no.5 s.94
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• pp.341-344
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• 2005

This research was performed for developing of biological treatment process of odor gas such as MEK, $H_2S$, and toluene, which was generated from the food waste recycling process. To establish the operational conditions of odor gas removal by small-scale biofiltration equipment, it was continuously operated by using toluene as a treating odor object. When the odor treating microorganisms were adhered to fibril form biofilter, high removal efficiency over $93\%$ was obtained by biofilm formation. At 400 ppm of inlet odor gas concentration and 10 sec of retention time, the removal efficiency was $76\%$ and $93\%$ in 1 st stage reactor and End stage reactor, respectively. However, the removal efficiency remained over $97\%$ at the operational conditions above 15 sec of retention time.

• Journal of Environmental Science International
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• v.20 no.3
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• pp.291-299
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• 2011

The main objectives of this research are to investigate characteristics of ozone solubility due to low solubility of conventional bubbles-ozone generators, evaluate the treatment characteristics of reclaiming textile wastewater for industrial water by means of micro/nano bubbles-dissolved ozone flotation(MNB-DOF) process. The textile wastewater used in this research was obtained from final effluent of the textile wastewater in B city. There is a 400L reactor which consists of a micro-nano bubble system and a ozone generator for experiments. As a result of generating micro-nano bubbles (below $0.5{\mu}m$) by using of MNB-DOF process, it improved ozone solubility due to higher ozone transfer rates. Consequently, the shorter ozonation time clearly indicates the lower power costs. The reported results clearly indicated that MNB-DOF process can be effectively and inexpensively. Results of the experiments through MNB-DOF process in this study satisfy all reclaiming standards as industrial water: pH 6.5~8.5, SS 10 mg/L or below, $BOD_5$) 6 mg/L or below, turbidity 10 NTU or below, Coliforms 1,000/100 mL or below. Therefore there is a possibility of the reclaiming of the textile wastewater as industrial water.

• Ecology and Resilient Infrastructure
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• v.4 no.3
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• pp.142-148
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• 2017

Nitrogen and phosphorus are key factors in causing eutrophication of water body. In this study, ceramics media was selected to increase the removal efficiency of nitrogen and phosphorus. We designed vertical, horizontal flow constructed wetlands to create aerobic and anaerobic flow conditions by using the media, then proceeded to performance evaluations after acrylic reactors were produced. In the case of vertical and horizontal flow constructed wetlands, we measured oxygen concentrations to evaluate aerobic and anaerobic conditions. we got the result of 2.7 mg/L in the aerobic condition, N.D in the anaerobic condition respectively, which suited our purpose. The result of the combined vertical and horizontal flow condition showed that the removal efficiency of SS was 94%, 91%, 61% at 140 min, 80 min, 60 min of running times, respectively, and the removal efficiency of T-P was 84%, 71%, 63% during each running time. In case of T-N, the removal efficiency was 63%, 49%, 42% during each running time. We found that the reactor exerted better removal efficiency when in the short time compared to 12 - 24 hr residence time of existing wetlands. In this study, we conducted experiments to explore functional effects after applying combined vertical and horizontal flow methods in the field. Further study will be carried out to identify its mechanism and administrative perspective.