• Proceedings of the KSRS Conference
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• 2006.03a
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• pp.178-186
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• 2006

기상 수치예보는 (Numerical Weather Pridiction, NWP)는 바람, 기온, 등과 같은 기상요소의 시간 변화를 나타내는 물리방정식을 컴퓨터로 풀어 미래의 대기 상태를 예상하는 과학적인 방법으로 지구를 상세한 격자 2진부호(GRIdded Binary, 이하 GRIB)로 나누어 그 격자점에서의 값으로 대기 상태를 나타낸다. 지구 각지에서의 각종 관측자료를 기초로 격자점상의 현재값을 구한다. 대용량의 격자데이터는 이진형태이어서 컴퓨터, 서버 저장장치에서 동일형태 데이터로 존재한다. 우리나라 최초의 저궤도 관측 위성인 다목적 실용위성 KOMPSAT-1호(이하, 아리랑 위성1호)는 전자광학카메라(Electro Optical Camera, EOC)를 탑재하여 1999년 12월 21일에 발사된 이후 2006년 1월 현재까지 6여년간 성공적으로 임무를 수행, 7049여회의 영상을 획득하여 국가적으로 귀중한 자료로 활용하고 있다. 아리랑 위성1호는 일일 2-3회 EOC영상을 획득하고 있으며, 임무계획(Mission Planning)은 MP(Mission Planner)가 사용자로부터 자료를 수집하여 임무분석 및 계획 서브시스템(MAPS)에 의해 계산되어진 위성의 제도예측 데이터에 촬영하고자하는 목표지점 좌표를 입력하여 자동명령생성기(KSCG)에 의해 계산된 촬영 경사각도(Tilt)값을 위성에 전송하여 목표지역의 영상을 획득하게 된다. 위성영상 획득에 있어 고가의 위성을 운영하면서 기상의 상태를 정확히 예측하여 실패없이 유효한 영상을 획득하는 것이 무엇보다 중요하다. 본 논문에서는 효율적인 위성임무계획을 위한 기상수치예보 자료를 분석하여 앞으로 발사하게 될 고해상 카메라 탑제위성인 아리랑 위성2호와 3호에 적용하고자 한다. the sufficient excess reactivity to override this poisoning must be inserted, or its concentration is decreased sufficiently when its temporary shutdown is required. As ratter of fact, these have an important influence not only on reactor safety but also on economic aspect in operation. Considering these points in this study, the shutdown process was cptimized using the Pontryagin's maximum principle so that the shutdown mirth[d was improved as to restart the reactor to its fulpower at any time, but the xenon concentration did not excess the constrained allowable value during and after shutdown, at the same time all the control actions were completed within minimum time from beginning of the shutdown.및 12.36%, $101{\sim}200$일의 경우 12.78% 및 12.44%, 201일 이상의 경우 13.17% 및 11.30%로 201일 이상의 유기의 경우에만 대조구와 삭제 구간에 유의적인(p<0.05) 차이를 나타내었다.는 담수(淡水)에서 10%o의 해수(海水)

• Journal of the Korean Society for Nondestructive Testing
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• v.33 no.5
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• pp.465-471
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• 2013

A steam generator in nuclear power plant is a heatexchager which is used to convert water into steam from heat produced in a nuclear reactor core, and the steam produced in steam generator is delivered to the turbine to generate electricity. Because of damage to steam generator tubing may impair its ability to adequately perform required safety functions in terms of both structural integrity and leakage integrity, eddy current testing is periodically performed to evaluate the integrity of tubes in steam generator. This assessment is normally performed during a reactor refueling outage. Currently, the eddy current testing for steam generator of nuclear power plant in Korea is performed in accordance with KEPIC & ASME Code requirements, the eddy current testing system is consists of remote data acquisition unit and data analysis program to evaluate the acquired data. The KEPIC & ASME Code require that the electrical properties of remote data acquisition unit, such as total harmonic distortion, input & output impedance, amplifier linearity & stability, phase linearity, bandwidth & demodulation filter response, analog-to-digital conversion, and channel crosstalk shall be measured in accordance with the KEPIC & ASME Code requirements. In this paper, the measurement requirements of electrical properties for eddy current testing instrument described in KEPIC & ASME Code are presented, and the measurement results of newly developed eddy current testing instrument by KHNP(Korea Hydro & Nuclear Power Co., LTD) are presented.

• Nuclear Engineering and Technology
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• v.38 no.2
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• pp.163-174
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• 2006

The $Ph{\acute{e}}bus$ FP project is an international reactor safety project. Its main objective is to study the release, transport and retention of fission products in a severe accident of a light water reactor (LWR). The FPT4 test was performed with a fuel debris bed geometry, to look at late phase core degradation and the releases of low volatile fission products and actinides. Post Test Analyses results indicate that releases of noble gases (Xe, Kr) and high-volatile fission products (Cs, I) were nearly complete and comparable to those obtained during $Ph{\acute{e}}bus$ tests performed with a fuel bundle geometry (FPT1, FPT2). Volatile fission products such as Mo, Te, Rb, Sb were released significantly as in previous tests. Ba integral release was greater than that observed during FPT1. Release of Ru was comparable to that observed during FPT1 and FPT2. As in other $Ph{\acute{e}}bus$ tests, the Ru distribution suggests Ru volatilization followed by fast redeposition in the fuelled section. The similar release fraction for all lanthanides and fuel elements suggests the released fuel particles deposited onto the plenum surfaces. A blockage by molten material induced a steam by-pass which may explain some of the low releases. The revaporisation testing under different atmospheres (pure steam, $H_2/N_2$ and steam /$H_2$) and up to $1000^{\circ}C$ was performed on samples from the first upper plenum. These showed high releases of Cs for all the atmospheres tested. However, different kinetics of revaporisation were observed depending on the gas composition and temperature. Besides Cs, significant revaporisations of other elements were observed: e.g. Ag under reducing conditions, Cd and Sn in steam-containing atmospheres. Revaporisation of small amounts of fuel was also observed in pure steam atmosphere.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.12
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• pp.1421-1426
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• 2014

To improve the inherent safety of the sodium-cooled fast reactor (SFR), the supercritical $CO_2$ ($S-CO_2$) Brayton cycle is being considered as an alternative power conversion system to steam the Rankine cycle. In the $S-CO_2$ system, a PCHE (printed circuit heat exchanger) is being considered. In this type of heat exchangers, diffusion bonding is used for joining the thin plates. In this study, the diffusion bonding characteristics of various austenitic alloys were evaluated. The tensile properties were measured at temperatures starting from the room temperature up to $650^{\circ}C$. For the 316H and 347H types of stainless steel, the tensile ductility was well maintained up to $550^{\circ}C$. However, the Incoloy 800HT showed lower strength and ductility at all temperatures. The microstructure near the bond line was examined to understand the reason for the loss of ductility at high temperatures.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.12
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• pp.1415-1420
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• 2014

Super-critical $CO_2$ ($S-CO_2$) Brayton cycle has been considered to replace the current steam Rankine cycle in Sodium-cooled Fast Reactor (SFR) in order to improve the inherent safety and thermal efficiency. Several austenitic alloys are considered as the structural materials for high temperature $S-CO_2$ environment.. Microstructural change after long-term exposure to high temperature $S-CO_2$ environment could affect to the mechanical properties. In this study, candidate materials (austenitic stainless steels and Alloy 800HT) were exposed to $S-CO_2$ to assess oxidation resistance and the change in tensile properties. Loss of ductility was observed for some austenitic stainless steels even after 250 h exposure. The contribution of $S-CO_2$ environment on such changes was analyzed based on the characterization of the surface oxide and carburization of the materials in which 316H and 800H showed different oxidation behaviors.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.36 no.5
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• pp.513-521
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• 2012

The aim of this study was to determine the brittle fracture behavior of reactor pressure vessel steel by considering the temperature dependence of a damage model. A multi-island genetic algorithm was linked to a Weibull stress model, which is the model typically used for brittle fracture evaluation, to improve the calibration procedure. The improved calibration procedure and fracture toughness test data for SA508 carbon steel at the temperatures $-60^{\circ}C$, $-80^{\circ}C$, and $-100^{\circ}C$ were used to decide the damage parameters required for the brittle fracture evaluation. The model was found to show temperature dependence, similar to the case of NUREG/CR-6930. Finally, on the basis of the quantification of the difference between 2- and 3-parameter Weibull stress models, an engineering equation that can help obtain more realistic fracture behavior by using the simpler 2-parameter Weibull stress model was proposed.

• Nuclear Engineering and Technology
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• v.27 no.5
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• pp.645-660
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• 1995

The loss of Residual Heat Removal System (RHRS) event during reduced inventory operation for the Korean Standard Nuclear Power Plants (KSNPPS) is simulated by RELAP5/MOD3 and RELAP5/MOD3.1 Tn cases are considered : Base case for an intact Reactor Coolant System (RCS) with no tent and a vent case for an open system. Comparative simulations of base case are peformed by RELAP5/MOD3 and RELAP5/MOD3. 1 computer codes. The results of too simulations are generally in good qualitative and quantitative agreement. However, since the results of RELAP5/MOD3 simulation reveals the deficiency of RELAP5/MOD3 wall heat model, the RELAP5/AOD3.1 computer code is used for the simulation of the vent case. The analysis result of base case show that two steam generators are insufficient to remove decay heat at one day after shutdown, where the RCS is closed. The RCS pressure increased continuously and reached the RCS temporary boundaries design pressure of 0.24 MPa around 4,000 seconds. In the vent case with a flow capacity equivalent to three times the capacity of Pressurizer Safety Valve (PSV), it is shown that the RCS Pressure does not reach 0.24 MPa and core uncovery does not occur until 10,000 seconds. The detailed discussions on the results of this study suggest the feasibility of RELAP5/AOD3.1 as an analysis tool for the simulation of the loss of RHRS event at reduced inventory operation. The results of this study also provide insight for the determination of proper vent capacity.

• Journal of Korea Water Resources Association
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• v.46 no.11
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• pp.1069-1078
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• 2013

The Bangudae Petroglyphs, national treasure No. 285 is located within submerged upper districts of Sayeon dam supplying the main residential water in Ulsan. Of the many ways for the reservation of Petroglyphs located the altitude at 53~57 m, the plan that we take it out of the water lowering the water level from 60 m to 52 m has been examined mainly in case of controlling artificially the water level of the dam. In this paper, we examined expected problems from the loss of dam function and the change of water quality from water deterioration caused by the water level control of the Sayeon dam. Using the model of Vollenweider and CSTR (Continuous Stirred Tank Reactor), we analyzed the density change of BOD and COD, representative water quality index and the TP and TN, the main reason of algae growth. The result showed that the density of COD lowered a little but the density of TP and TN went up over 130% when controlling the water level from 60 m to 52 m. These changes cause a serious algae problem and if doing the water quality management as the density of TN and TP, the water quality would become worse. Water storage and supply residential water decreases, and the water quality becomes worse because of eutrophic state.

• Proceedings of KOSOMES biannual meeting
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• 2007.05a
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• pp.91-96
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• 2007

Lab scale experimental study was carried out for SBR process, to investigate the effects of influent ship sewage organic compound removal and Bacillus sp. state on design parameters. This process was able to remove nitrogen and phosphorus as well as organic matter efficiently. More than 95% of chemical oxygen demand(COD) were removed. In addition, about 97% of total nitrogen (T-N) was reduced. The total phosphorus(T-P) reduction averaged 93%. The performance load of SBR process was shown to be $0.095kg{\cdot}TOC/m^3{\cdot}day$. The pH was decreased from 8.1 to 7.0 within 30 min and increased to 7.3 at the end of anoxic stage, and these phenomena were explained. The sludge produced in the SBR process is characterized by low generation rate (about $0.36kg{\cdot}MLSS/kg{\cdot}TOC$) and excellent settleability. The number of Bacillus sp. in the SBR was 24.2%, indicating that Bacillus sp. was a predominant species in the reactor.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.28 no.5
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• pp.535-542
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• 2015

A reactor containment acts as a final barrier to prevent leakage of radioactive material due to the possible reactor accidents into external environment. Because of the functional importance of the containment building, the SIT(Structural Integrity Test) for containments shall be performed to evaluate the structural acceptability and demonstrate the quality of construction. In this paper, numerical analyses are presented, which simulate the results obtained from the SIT for a prestressed concrete(PSC) structure. A sophisticate structural analysis model is developed to simulate the structural behavior during the SIT properly based on various preliminary analysis results considering contact condition among structural elements. From the comparison of the analysis and test results based on the acceptance criteria of ASME CC-6000, it can be concluded that the construction quality of the containment has been well maintained and the acceptable performance of new design features has been verified.