• 전기학회논문지P
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• 제59권3호
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• pp.325-329
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• 2010

Capacitor has been used principally for the power factor compensation long ago. However now it does as passive filter to reduce harmonics of nonlinear load with reactor. Most of the customer's low-voltage feeder has been designed with approximately balanced and connected at the 3 phase four wire system. But voltage and current unbalance is appeared by the mixing operation of single or three phase load etc. The addition of reactor at the condenser may rise its terminal voltage. Voltage and current values above rating can act on electrical stress on the condenser. In this paper, we calculated and measured that voltage, current and capacity of condenser are changed under the voltage balance. We conclude that magnitude and deviation of phase voltage act on major point of electrical stress.

• 상하수도학회지
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• 제13권4호
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• pp.35-44
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• 1999

In this study, it was attempted to remove nitrate and carbon in a single-stage reactor using acetate as substrate. Hybrid type upflow sludge baffled filter reactor was adopted using anaerobic sludge. Sludge bed in the bottom of reactor was intended to remove carbon and nitrate by denitrification and methanogenesis. And floating media in the upper part of reactor were intended to remove remaining carbon which was not removed due to the inhibition of nitrogen oxide on methane producing bacteria. The reactor removed over 96% of COD and most of nitrate with volumetric loading rate of $4.0kgCOD/m^3{\cdot}day$, hydraulic retention time of 24hr, 4,000mgCOD/L, and $266mgNO_3-N/L$. Nitrate in anaerobic sludge was converted to nitrogen gas(denitrification) or ammonia (ammonification) according to pH of influent, COD removal efficiency was easily affected by the change of volumetric loading rates and nitrate concentration. And when influent pH was about 4.7, most nitrate changed to ammonia while when influent pH was about 6.8~7.0, most nitrate denitrified independent of $COD/NO_3-N$ ratio. Most granules were gray and a few were black. In gray-colored granule, black inner side was covered with gray substance and SEM illustrated Methanoccoci type microorganisms which were compact spherical shape. Anaerobic filter removed residual COD effectively which was left in sludge bed due to the inhibition of nitrogen oxide.

• 한국안전학회지
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• 제32권2호
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• pp.147-152
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• 2017

The external reactor vessel cooling (ERVC) is well known strategy to mitigate a severe accident at which nuclear fuel inside the reactor vessel is molten. In order to compare the heat removal capacity of ERVC between the nuclear reactor designs quantitatively, numerical method is often used. However, the study for ERVC using computational fluid dynamics (CFD) is still quite scarce. As a validation study on the numerical prediction for ERVC using CFD, the subcooled boiling flow and natural circulation of coolant at the ULPU-V experiment was simulated. The commercially available CFD software ANSYS-CFX was used. Shear stress transport (SST) model and RPI model were used for turbulence closure and wall-boiling, respectively. The averaged flow velocities in the downcomer and the baffle entry under the reactor vessel lower plenum are in good agreement with the available experimental data and recent computational results. Steam generated from the heated wall condenses rapidly and coolant flows maintains single-phase flow until coolant boils again by flashing process due to the decrease of saturation temperature induced by higher elevation. Hence, the flow rate of coolant natural circulation does not vary significantly with the change of heat flux applied at the reactor vessel, which is also consistent with the previous literatures.

• 한국기계기술학회지
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• 제13권2호
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• pp.145-153
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• 2011

The irradiation tests of materials in HANARO have been performed usually at temperatures below $300^{\circ}C$ at which the RPV(Reactor Pressure Vessel) materials of the commercial reactors such as the light water reactor and CANDU are operated. As VHTR(Very High Temperature Reactor) and SFR(Sodium-cooled Fast Reactor) projects are being carried as a part of the present Gen-IV program in Korea, the requirements for irradiation of materials at temperatures higher than $500^{\circ}C$ are recently being gradually increased. To overcome the restriction in the use at high temperature of the existing Al thermal media, a new capsule with double thermal media composed of two kinds of materials such as Al-Ti and Al-graphite was designed and fabricated more advanced than the single thermal media capsule. At the irradiation test of the capsule, the temperature of the specimens successfully reached $700^{\circ}C$ and the integrity of Al, Ti and graphite material was maintained.

• Nuclear Engineering and Technology
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• 제25권2호
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• pp.276-284
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• 1993

ABB-CE사의 System-80 설계 특성 중 원자로 출력 급감발 제어계통(RPCS : Reactor Power Cutback System)은 2개의 주급수 펌프 중 1대가 정지하거나 전출력 부하 상실사고인 경우에도 원자로 정지없이 운전하게 함으로써 원전의 경제성 향상에 도움을 주고 있다. 이러한 RPCS의 적용 범위를 확대하여 단일제어봉 낙하를 포함한 제어봉 인입편차(inward deviation)가 발생하는 경우에도 RPCS를 작동시키면 원자로를 정지시키지 않고 운전을 계속할 수 있는지를 분석하였다. 즉 제어봉 인입편차가 발생시 제어봉을 순간적으로 낙하시켜 1차계통의 출력을 낮추면서 원자로를 정지시키지 않고도 과도현상을 수습할 수 있는지 분석하였다. 이렇게 확대된 RPCS는 미국 EPRI의 개량형 경수로 요건사항을 만족하는 것이며 제어봉 인입편차의 과도상태를 수용할 수 있도록 하는 ABB-CE사의 System-80+ 설계 항목에도 포함되어 있다. 본 연구에서는 System-8O+에 대하여 RPCS의 작동에 의한 제어봉의 삽입과 그에 따른 핵증기 공급계통의 변화를 모사할 수 있는 노심해석 모델을 개발하였다. 연구 결과 단일 제어봉 낙하를 포함한 제어봉 인입편차가 발생되어도 원자로 출력 급감발 제어를 확대 적용하는 경우 원자로 정지를 방지할 수 있게 되어 원전의 이용율을 향상시킬 수 있을 것으로 검토되었다.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1388-1395
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• 2017

The Helically coiled tube Once-Through Steam Generator (H-OTSG) is a key piece of equipment for compact small reactors. The present study developed and verified a thermal-hydraulic design and performance analysis computer code for a countercurrent H-OTSG installed in a small pressurized water reactor. The H-OTSG is represented by one characteristic tube in the model. The secondary side of the H-OTSG is divided into single-phase liquid region, nucleate boiling region, postdryout region, and single-phase vapor region. Different heat transfer correlations and pressure drop correlations are reviewed and applied. To benchmark the developed physical models and the computer code, H-OTSGs developed in Marine Reactor X and System-integrated Modular Advanced ReacTor are simulated by the code, and the results are compared with the design data. The overall characteristics of heat transfer area, temperature distributions, and pressure drops calculated by the code showed general agreement with the published data. The thermal-hydraulic characteristics of a typical countercurrent H-OTSG are analyzed. It is demonstrated that the code can be utilized for design and performance analysis of an H-OTSG.

• Nuclear Engineering and Technology
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• 제35권5호
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• pp.454-470
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• 2003

In this paper, a preliminary study for development of a fault diagnosis is presented for monitoring and diagnosing faults in the agitator driving system of a high temperature reduction reactor. In order to identify a fault occurrence and classify the fault cause, vibration signals measured by accelerometers on the outer shroud of the agitator driving system are firstly decomposed by wavelet transform (WT) and the features corresponding to each fault type are extracted. For the diagnosis, the fuzzy ARTMAP is employed and thereby, based on the features extracted from the WT, the robust fault classifier can be implemented with a very short training time - a single training epoch and a single learning iteration is sufficient for training the fault classifier. The test results demonstrate satisfactory classification for the faults pre-categorized from considerations of possible occurrence during experiments on a small-scale reduction reactor.

• 한국압력기기공학회 논문집
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• 제10권1호
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• pp.64-69
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• 2014

Gr. 91 steel is used for the major structural components of Generation-IV reactor systems, such as a very high temperature reactor(VHTR) and sodium-cooled fast reactor(SFR). Since these structures are designed for up to 60 years at elevated temperatures, the prediction of long-term creep life is important for a design application of Gr. 91 steel. In this study, a number of creep rupture data were collected through world-wide literature surveys, and using these data, the long-term creep life was predicted in terms of three methods: the single-C method in Larson-Miller(L-M) parameter, multi-C constant method in the L-M parameter, and a modified method("sinh" equation) in the L-M parameter. The results of the creep-life prediction were compared using the standard deviation of error value, respectively. Modified method proposed by the "sinh" equation revealed better agreement in creep life prediction than the single-C L-M method.

• Nuclear Engineering and Technology
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• 제52권3호
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• pp.538-543
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• 2020

Boric acid is used as a coolant for pressurized-water reactors, and the degree of burnup is controlled by the concentration of boric acid. Therefore, accurate measurement of the concentration of boric acid is an important factor in reactor safety. An improved system was proposed for the accurate determination of boron concentration. A new boron-concentration measurement technique, called multisensitive detection, was developed to improve the measurement accuracy of boron meters. In previous studies, laboratory-scale experiments were performed based on different sensitivity detectors, confirming a 65% better accuracy than conventional single-detector boron meters. Based on these experimental results, an experimental system simulating the coolant-circulation environment in the reactor was constructed; accuracy analysis of the boron meter with a multisensitivity detector was performed at the actual coolant pressure and temperature. In this study, the boron concentration conversion equation was derived from the calibration test, and the accuracy of the boron concentration conversion equation was examined through a repeatability test. Through the experiment, it was confirmed that the accuracy was up to 87.5% higher than the conventional single-detector boron meter.

• Nuclear Engineering and Technology
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• 제45권3호
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• pp.367-376
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• 2013

This study investigated mixing scenarios of the low and high conversion ratios (CRs) of fast reactors (FRs). The fuel cycle was modeled so as to minimize the spent fuel (SF) or transuranics (TRU) inventories. The scenarios were modeled for a single low CR of 0.61 and a high CR of 1.0. The study also investigated the mixing scenario of low-high CR and/or high-low CR. The SF and TRU inventories, associated with different scenarios, were compared to those of the light water reactor (LWR) once-through (OT) case. Also, the important isotope concentration and long-term heat (LTH) load were calculated and compared to those of the OT cycle. As a result, it is known that the deployment of FRs of low CR burns more TRU and results in a reduction of the out-of-pile TRU inventory and LTH with low deployment capacity. This study shows that the mixing strategy of FRs of low and high CR can reduce the SF and TRU inventories with lower deployment capacity as compared with a single deployment of FRs of high CR.