• Nuclear Engineering and Technology
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• 제56권7호
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• pp.2875-2880
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• 2024

The development of nuclear power plants is in three phases. The first phase is a consideration before the decision on the NPP construction program is approved, the second phase is the preparatory work for making contracts and preparing for the construction of NPP after the NPP construction policy is approved, and the third phase is contracting, licensing and building the first NPP. As a volcanically active country, Indonesia contains over 130 active volcanoes that are part of the Pacific Ring of Fire. The volcanic aspect is one of the safety factors considered while deciding the location of an NPP. Research on the potential of natural external risks to the determination of nuclear power plants in Indonesia, including the volcanic aspect, has been conducted based on the safety reference or safety guide of the IAEA and the Nuclear Energy Regulatory Body (BAPETEN) Regulation. Due to technological advancements, safety needs have evolved so the existing Indonesia National Standard (SNI) must be updated to comply with BAPETEN regulations. The substance in SNI 18-2034-1990 relating to volcanic features seems less relevant in actual conditions, given that more complete and exact criteria for determining a site guarantee the safety and health of residents and surrounding the environment site. The study intends to conduct a gap analysis of volcanic issues in SNI and volcanic regulations. The method used is identification requirements for volcanic aspects in SNI 18-2034-1990 about Determining Site of Nuclear Reactor Guidance with BAPETEN Chairman Regulation (BCR) number 4 of 2018 about Nuclear Installation Site Evaluation Safety Provisions and BCR number 5 of 2015 about Evaluation of Nuclear Installation Sites for Volcanic Aspects, and analysis uses a qualitative method of inductive techniques. The outcome of this research applies to suggesting a revision of SNI number 18-2034-1990, especially the volcanic aspect.

• 방사선산업학회지
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• 제5권2호
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• pp.185-189
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• 2011

As nuclear industry has been developed, a various types of radiological contamination has occurred. After 9.11 terror in U.S.A., it has been concerned that terrorists' active area has been enlarged to use nuclear or radioactive substance. Recently, the most powerful earth-quake stroke, which triggered a massive tsunami in Japan and then Fukushima nuclear power plant reactor has suffered from a serious accident in history. The Fukushima reactor accident has occurred an anxiety of radiation leaks and about 170,000 people have been evacuated from the accidental area near the nuclear power plant. For these reasons, a social chaos can be occurred if radiological contamination occurs to the supply system for the drinking water. As such, the establishment of the radiation monitoring system for the city main water system is compelling for the national security. In this study, a feasibility test of radiation monitoring system which consists of unified hybrid-type radiation detectors was experimented for multi detection system by using gamma-ray imaging. The hybrid-type radiation sensors were fabricated with CsI(Tl) scintillators and photodiodes. A preamplifier and amplifier was also fabricated and assembled with the sensor in the shielding case. For the preliminary test of detection of radiological contamination in the river, multi CsI(Tl)-PIN photodiode radiation detectors and $^{137}Cs$ gamma-ray source were used. The DAQ was done by Linux based ROOT program and NI DAQ system with Labview program. The simulated contamination was assumed to be occurred at Gapcheon river in Daejeon city. Multi CsI(Tl)-PIN photodiode radiation detectors were positioned at the Gapcheon river side. Assuming that the radiological contaminations flows in the river the $^{137}Cs$ gamma-ray source has been moved and then, the contamination region was reconstructed.

• 한국유체기계학회 논문집
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• 제17권6호
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• pp.104-108
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• 2014

The recent Fukushima nuclear power plant accidents shows that the core make up at high RCS pressure condition is very important to prevent core melting. The core make up flow at high pressure condition should be driven by gravity force or passive forces because the AC-powered safety features are not available during a Station Black Out (SBO) accident. The reactor Coolant System (RCS) mass inventory is continuously decreased by releasing steam through the pressurizer safety valves after reactor trip during a SBO accident. The core will be melted down within 2~3 hours without core make up action by active or passive mode. In the new design concept of a Hybrid Safety Injection Tank (Hybrid SIT) both for low and high RCS pressure conditions, the low pressure nitrogen gas serves as a charging pressure for a LBLOCA injection mode, while the PZR high pressure steam provides an equalizing pressure for a high pressure injection mode such as a SBO accident. After the pressure equalizing process by battery driven initiation valve at a high pressure SBO condition, the Hybrid SIT injection water will be passively injected into the reactor downcomer by gravity head. The SBO simulation by MARS code show that the core makeup injection flow through the Hybrid SIT continued up to the SIT empty condition, and the core heatup is delayed as much.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2009년도 제40회 하계학술대회
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• pp.638_639
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• 2009

The motors faults including mechanical rotor imbalances, broken rotor bar, bearing failure and eccentricities problems are reflected in electric, electromagnetic and mechanical quantities. This paper presents a study and the practical implementation of an induction motor for reactor containment fan cooler in nuclear power plant with Electric Signature Analysis(ESA). The results obtained present a good degree of reliability hence; the ESA predictive maintenance tools enable a pro-active evaluation of induction motors performance prior to failure.

• Nuclear Engineering and Technology
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• 제45권4호
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• pp.459-468
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• 2013

During a station blackout (SBO), the initiating event is a loss of Class IV and Class III power, causing the loss of the pumps, used in systems such as the primary heat transporting system (PHTS), moderator cooling, shield cooling, steam generator feed water, and re-circulating cooling water. The reference case of the SBO case does not credit any of these active heat sinks, but only relies on the passive heat sinks, particularly the initial water inventories of the PHTS, moderator, steam generator secondary side, end shields, and reactor vault. The reference analysis is followed by a series of sensitivity cases assuming certain system availabilities, in order to assess their mitigating effects. This paper also establishes the strategies to mitigate SBO accidents. Current studies and strategies use the computer code of the Integrated Severe Accident Analysis Code (ISAAC) for Wolsong plants. The analysis results demonstrate that appropriate strategies to mitigate SBO accidents are established and, in addition, the symptoms of the SBO processes are understood.

• Nuclear Engineering and Technology
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• 제39권2호
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• pp.103-110
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• 2007

The High Temperature Gas-cooled Reactor (HTGR) possesses inherent safety features and is recognized as a representative advanced nuclear system for the future. Based on the success of the HTR-10, the long-time operation test and safety demonstration tests were carried out. The long-time operation test verifies that the operation procedure and control method are appropriate for the HTR-10 and the safety demonstration test shows that the HTR-10 possesses inherent safety features with a great margin. Meanwhile, two new projects have been recently launched to further develop HTGR technology. One is a prototype modular plant, denoted as HTR-PM, to demonstrate the commercial capability of the HTGR power plant. The HTR-PM is designed as $2{\times}250$ MWt, pebble bed core with a steam turbine generator that serves as an energy conversion system. The other is a gas turbine generator system coupled with the HTR-10, denoted as HTR-10GT, built to demonstrate the feasibility of the HTGR gas turbine technology. The gas turbine generator system is designed in a single shaft configuration supported by active magnetic bearings (AMB). The HTR-10GT project is now in the stage of engineering design and component fabrication. R&D on the helium turbocompressor, a key component, and the key technology of AMB are in progress.

• Tribology and Lubricants
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• 제30권2호
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• pp.124-129
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• 2014

The piping system accounts for a large portion of the machinery structure of a plant, and is considered as a very important mechanical structure for plant safety. Accordingly, it is used in most energy plants in the nuclear, gas, and heavy chemical industries. In particular, the piping system for a nuclear plant is generally complicated and uses the reactor and its cooling system. The piping equipment is exposed to diverse loads such as weight, temperature, pressure, and seismic load from pipes and fluids, and is used to transfer steam, oil, and gas. In ultrasound infrared thermography, which is an active thermography technology, a 15-100 kHz ultrasound wave is applied to the subject, and the resulting heat from the defective parts is measured using a thermography camera. Because this technique can inspect a large area simultaneously and detect defects such as cracks and delamination in real time, it is used to detect defects in the new and renewable energy, car, and aerospace industries, and recently, in piping defect detection. In this study, ultrasound infrared thermography is used to detect information for the diagnosis of nuclear equipment and structures. Test specimens are prepared with piping materials for nuclear plants, and the optimally designed ultrasound horn and ultrasound vibration system is used to determine damages on nuclear plant piping and detect defects. Additionally, the detected images are used to improve the reliability of the surface and internal defect detection for nuclear piping materials, and their field applicability and reliability is verified.

• 자원리싸이클링
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• 제19권6호
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• pp.43-50
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• 2010

에너지 위기로 인하여 석탄에 대한 관심이 증가하고 있다. 그 중에서도 저등급석탄에 대한 관심이 증가하고 있는데, 저등급석탄은 수분함량이 30~60%갱도로 수분함량이 높다. 이러한 저등급 석탄을 발전용 연료로 사용하기 위해서는 건조공정이 선행적으로 이루어져야 한다. 본 연구에서는 고정층 반응기를 이용하여 저등급석탄의 건조 반응속도론을 도출하였다. 건조반응속도는 입자크기, 주입가스 온도, 가스 유속, L/D의 영향을 변수로 하여 도출하였다. Reynold's number는 가스 유속과 석탄업자의 크기, L/D는 반응기 직경과 대상탄의 충진양을 보정하기 위해 고려하였다. 석탄의 건조 특성에서도 알 수 있듯이, 고정층 반응기를 이용한 저등급석탄의 건조에 있어서도 표면수분의 건조가 원활하며, 상 경계 반응이 적합한 메커니즘임을 확인 할 수 있었다.

• 환경위생공학
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• 제11권2호
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• pp.43-52
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• 1996

Using rotating biological contactor(RBC) with artificial endogenous stage and aerobic fixed biofilm reactor(AFBR), organic material removal and biological nitrification of piggery wastewater has been studied at a pilot plant. RBC was operated in the endogenous phase at a interval of every 25 days. The concentration of COD, BOD and TKN in influent wastewater were from 2,940 to 3,800 mg/L, from 1,190 to 1,850 mg/L and from 486 to 754 mg/L respectively. The maximum active biomass content represented as VSS per unit aera was $2.0mg/cm$^{2}$ and biofilm dry density of $17mg/cm^{3}$ was observed at biofilm thickness of $900{\;}{\mu}m$. It was observed that the pilot scale RBC/AFBR process exhibited 72 percentage to 93 percentage of BOD removal, In order to obtain more than 90 percentage of BOD removal, the organic loading rate to the RBC/AFBR process should be maintained less than $0.09{\;}m^{3}/m^{2}{\cdot}day(125.9g{;\}BOD/m^{3}{\cdot}d$. The TKN removal efficiencies was from 45.5 to 90.9 percentage according to vary influent loading rate, It was estimated that the RBC/AFBR process consumed approximately 6.2 mg/L(as $CaCO_{3}$) of alkalinity per 1 mg/L of $NH_{3}$-N oxidized as the nitrification took piace.

• Nuclear Engineering and Technology
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• 제42권5호
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• pp.568-575
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• 2010

Flow Accelerated Corrosion is an active degradation mechanism of CANDU feeder. The tight bend downstream to Gray loc weld connection, close to reactor face, suffers significant wall thinning by FAC. Extensive in-service inspection of feeder wall thinning is very difficult because of the intense radiation field, complex geometry, and space restrictions. Development of a knowledge-based inspection program is important in order to guarantee that adequate wall thickness is maintained throughout the whole life of feeder. Research results and plant experiences are reviewed, and the plant inspection databases from Wolsong Units One to Four are analyzed in order to support developing such a knowledge-based inspection program. The initial thickness before wall thinning is highly non-uniform because of bending during manufacturing stage, and the thinning rate is non-uniform because of the mass transfer coefficient distributed non-uniformly depending on local hydraulics. It is obvious that the knowledge-based feeder inspection program should focus on both fastest thinning locations and thinnest locations. The feeder wall thinning rate is found to be correlated proportionately with QV of each channel. A statistical model is proposed to assess the remaining life of each feeder using the QV correlation and the measured thicknesses. W-1 feeder suffered significant thinning so that the shortest remaining life barely exceeded one year at the end of operation before replacement. W-2 feeder showed far slower thinning than W-1 feeder despite the faster coolant flow. It is believed that slower thinning in W-2 is because of higher chromium content in the carbon steel feeder material. The average Cr content of W-2 feeder is 0.051%, while that value is 0.02% for W-1 feeder. It is to be noted that FAC is reduced substantially even though the Cr content of W-2 feeder is still very low.