• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.52-57
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• 1998

In this paper, we performed a design study of a thorium fueled reactor according to the design concept of the Radkowsky Thorium Reactor (RTR) and evaluated its overall performance. To enhance its performance and alleviate its problems, we introduced a new metallic uranium fuel, uranium-zirconium hydride (U-Zr $H_{1.6}$), as a seed fuel. For comparison, typical ABB/CE-type PWR based on SYSTBM 80+ and standard RTR-type thorium reactor were also studied. From the results of performance analysis, we could ascertain advantages of RTR-type thorium fueled reactor in proliferation resistance, fuel cycle economics, and back-end fuel cycle. Also, we found that enhancement of proliferation resistance and safer operating conditions may be achieved by using the U-Zr $H_{l.6}$ fuel in the seed region without additional penalties in comparison with the standard RTR's U-Zr fuelr fuelel

• Nuclear Engineering and Technology
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• 제56권2호
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• pp.728-744
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• 2024

The presence of hydrogen absorbed by zirconium-based cladding materials during reactor operation can trigger degradation mechanisms and endanger the rod integrity. Ensuring the durability of the rods in extended time-frames like dry storage requires anticipating hydrogen behavior using numerical modeling. In this context, the present paper describes a hydrogen post-processing tool for Falcon - HYPE, a PSI's in-house tool able to calculate hydrogen uptake, transport, thermochemistry, reorientation of hydrides and hydrogen-related failure criteria. The tool extracts all necessary data from a Falcon output file; therefore, it can be considered loosely coupled to Falcon. HYPE has been successfully validated against experimental data and applied to reactor operation and interim storage scenarios to present its capabilities.

• 한국진공학회:학술대회논문집
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• 한국진공학회 2010년도 제39회 하계학술대회 초록집
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• pp.314-314
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• 2010

For the optimal design of a tokamak-type reactor, self-consistent determination of a radial build of reactor systems is important and the radial build has to be determined by considering the plasma physics and engineering constraints which inter-relate various reactor systems. In a low aspect ratio (LAR) tokamak reactor with a superconducting toroidal field (TF) coil, the shield should provide sufficient protection for the superconducting TF coil and the shield plays a key role in determining the size of a reactor. To determine the radial build of a reactor, neutronic effects such as tritium breeding in the blanket, nuclear heating, and radiation damage to toroidal field (TF) coil has to be included in the systems analysis. In this work, the outboard blanket only is considered where tritium self-sufficiency is possible by using an inboard neutron reflector instead of breeding blanket. The reflecting shield should provide not only protection for the superconducting TF coil but also improved neutron economy for the tritium breeding in outboard blanket. Tungsten carbide, metal hydride such as titanium hydride and zirconium hydride can be used for improved shielding performance and thus smaller shield thickness. With the use of advanced technology in the shield, conceptual design of a compact superconducting LAR reactor with aspect ratio of less than 2 will be presented as a viable power plant.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 추계학술대회논문집A
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• pp.843-848
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• 2000

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and it's containment vessel. If a flaw is found during the periodic inspection from the pressure tubes, the integrity evaluation must be carried out, and the safety requirements must be satisfied for continued service. In order to complete the integrity evaluation, complicated and iterative calculation procedures are required. Besides, a lot of data and knowledge for the evaluation are required for the entire integrity evaluation process. For this reason, an integrity evaluation system, which provides efficient way of evaluation with the help of attached databases, was developed. The developed system was built on the basis of ASME Sec. XI and FFSG(Fitness For Service Guidelines for zirconium alloy pressure tubes in operating CANDU reactors) issued by the AECL, and covers the delayed hydride cracking(DHC). Various analysis methods are provided for the integrity evaluation of pressure tube. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• 한국수소및신에너지학회논문집
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• 제24권2호
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• pp.128-135
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• 2013

The nuclear fuel cycle plant is composed of various subsystems such as a fuel storage and delivery system (SDS), a tokamak exhaust processing system, a hydrogen isotope separation system, and a tritium plant analytical system. Korea is sharing in the construction of the International Thermonuclear Experimental Reactor (ITER) fuel cycle plant with the EU, Japan, and the US, and is responsible for the development and supply of the SDS. Hydrogen isotopes are the main fuel for nuclear fusion reactors. Metal hydrides offer a safe and convenient method for hydrogen isotope storage. The storage of hydrogen isotopes is carried out by absorption and desorption in a metal hydride bed. These reactions require heat removal and supply respectively. Accordingly, the rapid storage and delivery of hydrogen isotopes are enabled by a rapid cooling and heating of the metal hydride bed. In this study, we designed and manufactured a vertical-type hydrogen isotope storage bed, which is used to enhance the cooling performance. We present the experimental details of the cooling performances of the bed using various cooling parameters. We also present the modeling results to estimate the heat transport phenomena. We compared the cooling performance of the bed by testing different cooling modes, such as an isolation mode, a natural convection mode, and an outer jacket helium circulation mode. We found that helium circulation mode is the most effective which was confirmed in our model calculations. Thus we can expect a more efficient bed design by employing a forced helium circulation method for new beds.

• 대한기계학회논문집A
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• 제24권1호
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• pp.206-214
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• 2000

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and it's containment vessel. If a flaw is found during the periodic inspection from the pressure tube s. the integrity evaluation must be carried out. and the safety requirements must be satisfied for continued service. In order to complete the integrity evaluation, complicated and iterative calculation procedures are required. Besides, a lot of data and knowledge for the evaluation are required for the entire: integrity evaluation process. For this reason. an integrity evaluation system, which provides efficient of evaluation with the help of attached databases, was developed. The developed system was built on the basis of ASME Sec. XI and FFSG(Fitness For Service Guidelines for zirconium alloy pressure tubes in operating CANDU reactors) issued by the AECL, and covers the delayed hydride cracking(DHC). This system does not only provide various databases including the 3-D finite element analysis results on pressure tubes, inspection data and design specifications but also is compatible with other commercial database software. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 추계학술대회
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• pp.1133-1138
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• 2003

The pressure tubes, which contain high temperature heavy water and fuel, are within the core of a CANDU nuclear reactor, and are thus subjected to high stresses, temperature gradient, and neutron flux. Further, it is well known that pressure tubes of cold-worked Zr-2.5Nb materials result in hydrogen diffusion, which create fully-hydrided regions (frequently called Blister). Thus a proper investigation of hydrogen diffusion within zirconium-alloy nuclear components, such as CANDU pressure tube and fuel channels is essential to predict the structural integrity of these components. In this respect, this paper presents numerical investigation of hydrogen diffusion to quantify the hydrogen concentration for blister growth of CANDU pressure tube. For this purpose, coupled temperature-hydrogen diffusion analyses are performed by means of two-dimensional finite element analysis. Comparison of predicted temperature field and blister with published test data shows good agreement.

• Nuclear Engineering and Technology
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• 제48권4호
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• pp.893-905
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• 2016

Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

• 한국추진공학회:학술대회논문집
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• 한국추진공학회 2017년도 제48회 춘계학술대회논문집
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• pp.451-455
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• 2017

ZPP와 THPP 화약의 압력 카트리지가 밀폐용기에 장착되어 연소될 때의 현상을 반응성 오일러리안-라그랑지안 이상 유동 해석 코드를 통해 모사 하였다. ZPP와 THPP는 압력 카트리지 내에서 boron nitride 판으로 격리되어있고, ZPP만 열선에 의해 직접 점화되기 때문에 THPP의 연소지연효과가 발생할 가능성이 높다. 실험을 통한 THPP의 점화지연 측정은 힘들기 때문에 기존의 연구를 통해 검증된 수치해석 코드를 통해 점화지연에 대한 케이스 스터디를 수행하고 현상학적 분석을 수행하였다. 해석 결과 THPP의 점화지연 정도에 따라 초기 충격파의 강도가 변하여 압력선도의 초기 피크특성 뿐만 아니라 주파수에도 영향을 미친다는 것을 알 수 있었다.

• 한국재료학회지
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• 제27권12호
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• pp.672-678
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• 2017

In the production of zirconium cladding tube, a pickling acid solution is used to remove surface contaminants, which generates tons of pickling acid waste. The waste pickling solution is a valuable resource of Hf-free Zr. Many studies have investigated separating the Hf-free Zr source from the waste pickling acid. The results showed that $Ba_2ZrF_8$ precipitates prepared from the waste pickling acid were useful as an electrolyte for the electrorefining of Zr in molten salt. In the present work, electrorefining was performed in a $Ba_2ZrF_8-LiF$ binary electrolyte to recover Zr from a Hf-free CuZr ingot anode prepared by electroreduction. Before electrorefining, two pretreatments are performed. First, electrolyte melting was carried out to determine the eutectic temperature, and second, the electrolyte was treated to eliminate impurities, mainly hydride. After electrorefining, the cathode deposits were analyzed by $O_2$ gas analyzer and SEM-EDX to explore the possibility of recovering nuclear-grade Zr metal. Moreover, the anode was analyzed by SEM-EDX to determine the Zr dissolution depth.