• Proceedings of the Korean Society of Precision Engineering Conference
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• 2001.04a
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• pp.651-654
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• 2001

In the dismantling process of nuclear spent fuels, the spent fuel rod cutting process, followed immediately by the decladding process, performs the cutting the spent fuel rods to a proper length for fast decladding operation. In this paper, we analyzed the chemical compositions, mechanical properties, and physical characteristics for SUS and zircaloy tubes in order to identify the feasibility of cutter-blade type in cutting SUS and zircaloy tubes. It is considered that material, shape and angle, and heat treatment for fabricating the highly durable cutter blade and also it is investigated that the round-shape sustenance of cross-section, amount of debris production, and fire occurrence for measuring the cutting performance on SUS and zircaloy tubes, spent fuel rod cutting device is designed to be operated automatically through the remote control system for use in Hot Cell(radioactive) area and the electro-driven mechanical parts are modularized for easy maintenance. Results from various experiments confirm the efficiency of this device.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.55-60
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• 1995

CANDU형 원자로의 노심해석을 위해 핵연료 격자 및 반응도 설비(reactivity devices)에 대한 2군 군정수가 필요하다. 특히 CANDU형 원자로의 노심해석에 있어서 반응도 설비나 구조물은 증분격자 상수(Incremental Cross Section)에 의해 묘사된다. 현재 CANDU형 원자로의 반응도 설비의 증분격자 상수를 계산하기 위해 MULTICELL 코드를 사용하여 계산하고 있다. 그러나 weak absorber에 대해 기존의 증분격자 상수를 이용하여 계산한 반응도가는 시운전(Phase-B)조건에서의 노물리 시험치보다 다소 과소평가하고 있다. 본 연구에서는 증분격자 상수 계산 방법의 개선 방향을 모색하기 위해 SHETAN 및 MCNP 코드로 단일 격자에서의 반응도가를 계산하여 비교, 평가하였다. HCNP 계산의 결과는 조정봉(Adjuster rods)과 흡수봉/정지봉 (Mechanical Control Absorber/Shutoff rod)은 MULTICELL의 계산 결과보다 적으며, 경수영역 조절기(Liquid Zone Controller)는 크게 나타났다. 또한 SHETAN 코드를 이용한 결과는 MULTICELL의 결과보다 약간 크게 나타났다.

• Journal of Biosystems Engineering
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• v.25 no.1
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• pp.19-24
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• 2000

This study was carried out to develop gripper which to adaptive variable tray and to develop automatic transplanting system for seedling-production system between tray. This system consisted of five set of gripper and end-effector, a planting-width control unit, a tray transfer unit, and gripper moving device which move gripper between nursing tray and growing tray. This system used push-out rod to grasp plant instead of pull-out end -effector. Several types of fingers, which physically grip seedlings, were also developed and tested to ensure reliable transplanting operation of the gripper. The transplanting system detaches seedlings from a tray with push-o0ut rods, which were installed under the tray transfer unit. The performance of the transplanting system was evaluated by successive transplanting experiments. Using the best type of finger , the transplanting system produced 94.6% of transplanting success rate.

• Nuclear Engineering and Technology
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• v.49 no.8
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• pp.1696-1710
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• 2017

An improved Reactor Regulating System (RRS) logic architecture, which is combined with genetic algorithm (GA), is implemented in this work. It is devised to provide an optimal solution to the current RRS. The current system works desirably and has contributed to safe and stable nuclear power plant operation. However, during the ascent and descent section of the reactor power, the RRS output reveals a relatively high steady-state error, and the output also carries a considerable level of overshoot. In an attempt to consolidate conservatism and minimize the error, this work proposes to apply GA to RRS and suggests reconfiguring the system. Prior to the use of GA, reverse engineering is implemented to build a Simulink-based RRS model. Reengineering is followed to produce a newly configured RRS to generate an output that has a reduced steady-state error and diminished overshoot level. A full-scope APR1400 simulator is used to examine the dynamic behaviors of RRS and to build the RRS Simulink model.

• Proceedings of the Korean Society for Technology of Plasticity Conference
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• 2008.05a
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• pp.482-484
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• 2008

Anodic aluminum oxide (AAO) which prepared with two-step anodizing method (with dissimilar solutions) was used as a template to fabricate highly ordered, free standing metal nano-rods. AAO nano-template technique can realize self-organized hexagonal pore structure with nanometer dimension size, it's easy to control pore diameter, length and density by varying anodizing conditions. Ni and Ni/Fe/Cu multi-metal layer nanorods were electrochemically deposited into AAO nano-template by AC voltage in simple sulfate solutions.. The properties of samples are tested by X-ray diffraction (XRD), field emission microscopy (FE-SEM).

• Proceedings of the Korean Society for Technology of Plasticity Conference
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• 2006.05a
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• pp.141-144
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• 2006

In this study, we developed a decladding device which separates 250 mm length of simulated nuclear spent fuel rod into the pallets and the pieces of the hulls after inserting the rod cut into the module with several pairs of blades. To improve the performance of the equipment, we considered some mechanisms to prevent the rod cut from being exposed or bounced into the hot-cell, to reduce the operation time, and to insert the rods automatically. It is expected that the newly developed system will contribute to prevent radioactive pollution in the hot-cell, reduce the operation time, and to increase the safety of the operators. As a result of the performance test for some mockup fuel rod cuts in the ACP(Advanced Spent Fuel Control Process) facility, it was verified that the decladding device could be applied to the actual fuel rod cut. And it will be able to use for a scale-up facility in the future.

• Particle and aerosol research
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• v.8 no.1
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• pp.37-43
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• 2012

In this study, we designed a 'spark in liquid' system. The spark discharge between two electrodes were used to generate particles by using sufficient temperature to evaporate a part of electrodes. The power supply system provides a continuous spark discharge by discharging of the capacitor to ionize the electrodes in liquid. The DC spark discharge system operates with 1-10 kV voltage. Processed copper and graphite rods were used to both electrodes with 1-3 mm diameter. There are several variables which can control the particle size and concentration such as gap distance between electrodes, applied voltage, operating liquid temperature, electrode type and liquid type. So we controlled these variables to confirm the change of particle size distribution and concentration of particles contained in liquid as wt%. 'spark in liquid' system is expected to apply nanoink by control of concentration with analysis of characteristics.

• Nuclear Engineering and Technology
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• v.46 no.4
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• pp.513-520
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• 2014

In this paper, three types of operational and industrial absorbers used at research reactors, including Ag-In-Cd alloy, $B_4C$, and Hf are selected for sensitivity analyses. Their integral effects on the main neutronic core parameters important to safety issues are investigated. These parameters are core excess reactivity, shutdown margin, total reactivity worth of control rods, thermal neutron flux, power density distribution, and Power Peaking Factor (PPF). The IAEA 10 MW benchmark core is selected as the case study to verify calculations. A two-dimensional, three-group diffusion model is selected for core calculations. The well-known WIMS-D4 and CITATION reactor codes are used to carry out these calculations. It is found that the largest shutdown margin is gained using the $B_4C$; also the lowest PPF is gained using the Ag-In-Cd alloy. The maximum point power densities belong to the inside fuel regions surrounding the central flux trap (irradiation position), surrounded by control fuel elements, and the peripheral fuel elements beside the graphite reflectors. The greatest and least fluctuation of the point power densities are gained by using $B_4C$ and Ag-In-Cd alloy, respectively.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.741-751
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• 2021

There has been a deep interest in trying to find better-performing fuel clad motivated by the desire to decrease the likelihood of the reactor barrier failure like what happened in Fukushima in recent years. In this study, the effect of move towards accident tolerant fuel (ATF) cladding as the most attracting concept for improving reactor safety is investigated for SMART modular reactor. These reactors have less production cost, short construction time, better safety and higher power density. The SiC and FeCrAl materials are considered as the most potential candidate for ATF cladding, and the results are compared with Zircaloy cladding material from reactor physics point of view. In this paper, the calculations are performed by generating PMAX library by DRAGON lattice physics code to be used for further reactor core analysis by PARCS code. The differential and integral worth of control and safety rods, reactivity coefficient, power and temperature distributions, and boric acid concentration during the cycle are analyzed and compared from the conventional fuel cladding. The rod ejection accident (REA) is also performed to study how the power changed in response to presence of the ATF cladding in the reactor core. The key quantitative finding can be summarized as: 20 ℃ (3%) decrease in average fuel temperature, 33 pcm (3%) increase in integral rod worth and cycle length, 1.26 pcm/℃ (50%) and 1.05 pcm/℃ (16%) increase in reactivity coefficient of fuel and moderator, respectively.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.781-789
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• 2022

KHNP had licensed Dynamic Control rod Reactivity Measurement (DCRM) method using detector current signals of PWRs in 2006. The method has been applied to all PWRs in Korea for about 15 years successfully. However, the original method was inapplicable to PWRs using low-sensitivity integral fission chamber as ex-core detectors because of their pulse pile-up and the nonlinearity of the mean-square voltage at low power region. Therefore, to overcome this disadvantage, a modified method, DCRM-EK, was developed using kinetics behavior after equilibrium condition where the pulse counts maintain the maximum value before pulse pile-up. Overall measurement, analysis procedure, and related computer codes were changed slightly to reflect the site test condition. The new method was applied to a total of 15 control rods of 1000 MWe and 1400 MWe PWRs in Korea with worths in the range of 200 pcm -1200 pcm. The results show the average difference of -0.4% and the maximum difference of 7.1% compared to the design values. Therefore, the new DCRM-EK will be applied to PWRs using low sensitivity integral fission chambers, and also can replace the original DCRM when the evaluation fails by big noises present in current or voltage signals of uncompensated/compensated ion chambers.