• 한국학교수학회논문집
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• 제25권4호
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• pp.333-352
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• 2022

본고는 초등수학을 기반으로 첨단기술및 ICT교구를활용한 초등융합교육자료의 개발에 대한 사례 연구이다. 본 연구에서는 초등현장 교육전문가 6인이 준비-개발-개선의 PDI 모형에 따라 융합교육 자료를 개발하였고, 개발된 자료는 수학, 과학, 미술, 공학 과목의 교과교육전문가 4인에 의해서 수정·보완하였다. 개발된 융합교육 자료는 초등 3~4학년군에서 활용 가능한 『그래프! 과거, 현재, 미래를 이어주는 다리』 의 1종이고, 5~6학년군은『알쏭달쏭, 같은 듯 다른 너!』 , 『함께 만드는 가상현실 입체 공간』의 2종이며, 초등 3~4학년군과 5~6학년군에서 학년군별로 재구성해 사용 가능한 『그리고 만드는 재미있는 도형 나라』 , 『거북선 지붕을 빈틈없이 덮어라!』 의 2종이었다. 본 개발 연구의 결과를 기반으로 한 제언은 다음과 같다. 초등교육 현장에서 원활한 융합교육을 수행할 수 있도록 다양한 융합교육 자료가 개발 공유되어야 하며, 나아가 앞으로 융합교육의 확산과 정착을 위해서는 현장 초등 교사나 예비 초등 교사들의 융합교육 역량의 함양 및 신장이 절대적인 선결 조건일 것이다.

• Nuclear Engineering and Technology
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• 제45권3호
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• pp.347-360
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• 2013

This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA) and the loss-of-feedwater accident (LOFW) in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF), a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.1052-1065
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• 2024

Advanced Pressurized Water Reactors (APWRs) and Boiling Water Reactors (BWRs) employ a suppression pool as a heat sink to prevent containment overpressure. Steam can be discharged into the pool through multi-hole spargers or blowdown pipes in both normal and accident conditions. Direct Contact Condensation (DCC) creates sources of momentum and heat. The competition between these two sources determines the development of thermal stratification or mixing of the pool. Thermal stratification is of safety concern as it reduces the cooling capability compared to a completely mixed pool condition. In this work we develop a scaling approach to prediction of the thermal stratification in a water pool induced by steam injection through spargers. Experimental data obtained from large-scale pool tests conducted in the PPOOLEX and PANDA facilities, as well as simulation results obtained using validated codes are used to develop the scaling. Two injection orientations, namely radial injection through multi-hole Sparger Head (SH) and vertical injection through Load Reduction Ring (LRR), are considered. We show that the erosion rate of the cold layer can be estimated using the Richardson number. In this work, scaling laws are proposed to estimate both the (i) transient erosion velocity and (ii) the stable position of the thermocline. These scaling laws are then implemented into a 1D model to simulate the thermal behavior of the pool during steam injection through the sparger.

• KEPCO Journal on Electric Power and Energy
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• 제2권3호
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• pp.425-431
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• 2016

Supply Capability of the generator, if the maximum demand occurs, refers to the maximum power that can be stably supplied and it is possible to maintain stable power supply to be greater than actual load. However, unexpected power demand and reduction in supply Capability due to stop of unexpected generator in operation can temporarily make a big chaos in power system. In fact, due to a lack of power supply Capability in the country, enforced emergency load adjustment to the September 15, 2011, the circulation power outage has occurred in several cities. As the result, interrupted operation of the elevator and stopped hospital medical equipment led to a great deal of trouble to people's lives, causing a social problem. At that time, it was found that a failed frequency control because of smaller actual supply Capability than that of predicted. The difference was about 1,170 MW with Gas turbine power plant. By accurately calculating the generator supply capability, we can not only grasp the power reserve rate, but also correspond to the time of power supply instability.

• Nuclear Engineering and Technology
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• 제41권5호
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• pp.617-648
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• 2009

Under the government supported long-term nuclear R&D program, the severe accident research program at KAERI is directed to investigate unresolved severe accident issues such as core debris coolability, steam explosions, and hydrogen combustion both experimentally and numerically. Extensive studies have been performed to evaluate the in-vessel retention of core debris through external reactor vessel cooling concept for APR1400 as a severe accident management strategy. Additionally, an improvement of the insulator design outside the vessel was investigated. To address steam explosions, a series of experiments using a prototypic material was performed in the TROI facility. Major parameters such as material composition and void fraction as well as the relevant physics affecting the energetics of steam explosions were investigated. For hydrogen control in Korean nuclear power plants, evaluation of the hydrogen concentration and the possibility of deflagration-to-detonation transition occurrence in the containment using three-dimensional analysis code, GASFLOW, were performed. Finally, the integrated severe accident analysis code, MIDAS, has been developed for domestication based on MELCOR. The data transfer scheme using pointers was restructured with the modules and the derived-type direct variables using FORTRAN90. New models were implemented to extend the capability of MIDAS.

• Nuclear Engineering and Technology
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• 제46권1호
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• pp.39-46
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• 2014

A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV) become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS) and the steam generator (SG) secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident.

• Nuclear Engineering and Technology
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• 제37권5호
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• pp.469-478
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• 2005

The subchannel analysis capability of MARS, a multi-dimensional thermal-hydraulic system code, has been enhanced. In particular, the turbulent mixing and void drift models for the flow-mixing phenomena in rod bundles were improved. Then, the subchannel analysis feature was combined with the existing coupled system thermal-hydraulics (T/H) and 3D reactor kinetics calculation capability of MARS. These features allow for more realistic simulations of both the hot channel behavior and the global system T/H behavior. Using the coupled features of MARS, a coupled analysis of a main steam line break (MSLB) is carried out for demonstration purposes. The results of the calculations are very reasonable and promising.

• Nuclear Engineering and Technology
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• 제46권6호
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• pp.869-874
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• 2014

It was discovered in a Korean PWR that an extensive number of very short and shallow cracks in the SG tubes were undetectable by eddy current in-service-inspection because of the masking effect of sludge deposits. Axial stress corrosion cracks at the outside diameter of the steam generator tubes near the line contacts with the tube support plates are the major concern among the six identical Korean nuclear power plants having CE-type steam generators with Alloy 600 high temperature mill annealed tubes, HU3&4 and HB3~6. The tubes in HB3&4 have a less susceptible microstructure so that the onset of ODSCC was substantially delayed compared to HU3&4 whose tubes are most susceptible to ODSCC among the six units. The numbers of cracks detected by the eddy current inspection jumped drastically after the steam generators of HB4 were chemically cleaned. The purpose of the chemical cleaning was to mitigate stress corrosion cracking by removing the heavy sludge deposit, since a corrosive environment is formed in the occluded region under the sludge deposit. SGCC also enhances the detection capability of the eddy current inspection at the same time. Measurement of the size of each crack using the motorized rotating pancake coil probe indicated that the cracks in HB4 were shorter and substantially shallower than the cracks in HU3&4. It is believed that the cracks were shorter and shallower because the microstructure of the tubes in HB4 is less susceptible to ODSCC. It was readily understood from the size distribution of the cracks and the quantitative information available on the probability of detection that most cracks in HB4 had been undetected until the steam generators were chemically cleaned.

• Journal of the Korean Wood Science and Technology
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• 제39권4호
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• pp.326-332
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• 2011

국산 산공재 중 구조적 특징이 목재흡음에 적합하다고 생각되는 백합나무 목재 횡단면의 흡음성능과 기체투과성의 방사방향 변동과 폭쇄처리 영향을 관찰하고자 무처리와 폭쇄처리 목재원반에서 방사방향위치가 다른 원형시험편을 채취하여 전달함수법과 CFP (capillary flow porometry)법으로 흡음율과 기체투과성을 각각 측정, 비교하였다. 측정주파수범위에서 폭쇄처리 횡단면의 흡음율이 무처리재보다 높은 흡음성능을 나타내었으며 횡단면에서는 대경 도관이 다수 존재하여 다공질형흡음에 유용한 연속된 모세관이 다량 존재하는 사실을 확인할 수 있었다. 방사방향으로는 심재부위에서 채취한 시험편보다 변재부위에서 채취한 시험편의 흡음계수가 높은 흡음율을 나타내었으며, 기체투과성도 변재부위가 심재부위보다 높은 수치를 나타내었다.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2699-2708
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• 2020

Helical-coil steam generator (HCSG) technology is a major design candidate for small modular reactors due to its compactness and capability to produce superheated steam with high generation efficiency. In this paper, we investigate the feasibility of the passively autonomous power maneuvering by coupling the 3-D transient multi-physics of a soluble-boron-free (SBF) core with a time-dependent HCSG model. The predictor corrector quasi-static method was used to reduce the cost of the transient 3-D neutronic solution. In the numerical system simulations, the feedwater flow rate to the secondary of the HCSGs is adjusted to extract the demanded power from the primary loop. This varies the coolant temperature at the inlet of the SBF core, which governs the passively autonomous power maneuvering due to the strongly negative coolant reactivity feedback. Here, we simulate a 100-50-100 load-follow operation with a 5%/minute power ramping speed to investigate the feasibility of the passively autonomous load-follow in a 450 MW_th SBF PWR. In addition, the passively autonomous frequency control operation is investigated. The various system models are coupled, and they are solved by an in-house Fortran-95 code. The results of this work demonstrate constant steam temperature in the secondary side and limited variation of the primary coolant temperature. Meanwhile, the variations of the core axial shape index and the core power peaking are sufficiently small.