• Nuclear Engineering and Technology
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• 제48권2호
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• pp.572-594
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• 2016

As one of the Gen-IV nuclear energy systems, a sodium-cooled fast reactor (SFR) is being developed at the Korea Atomic Energy Research Institute. As a long-term national research project, advanced radiation resistant oxide dispersion strengthened steel (ARROS) is being developed as an in-core fuel cladding tube material for a SFR in the future. In this paper, the current status of ARROS development is reviewed and its future prospective is discussed.

• 방사성폐기물학회지
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• 제21권1호
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• pp.115-147
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• 2023

The Fukushima accident in 2011 revealed some major flaws in traditional nuclear fuel materials under accidental conditions. Thus, the focus of research has shifted toward "accident tolerant fuel" (ATF). The aim of this approach is to develop fuel material solutions that lead to improved reactor safety. The application of protective coatings on the surface of nuclear fuel cladding has been proposed as a near-term solution within the ATF framework. Many coating materials are being developed and evaluated. In this article, an overview of different zirconium-based alloys currently in use in the nuclear industry is provided, and their performances in normal and accidental conditions are discussed. Coating materials proposed by different institutions and organizations, their performances under different conditions simulating nuclear reactor environments are reviewed. The strengths and weaknesses of these coatings are highlighted, and the challenges addressed by different studies are summarized, providing a basis for future research. Finally, technologies and methods used to synthesize thin-film coatings are outlined.

• 한국실내디자인학회논문집
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• 제14권6호
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• pp.228-235
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• 2005

This study analyzes the new possibility of design through the research for the expression of cladding design architectural materials, and applies to connection with creative design concept. In these modern days, architectural materials became the subject of expression as itself thanks to the development of new technology and materials. The physical properties of materials and their beauty value made possible the variety and possibility of expression in space design. The expressional properties of materials can lead mutual communication between commodity and brand by association, presenting brand image and the distinction of its value. Therefore, the trends of designed material expression in space design according to modern aspect can offer the possibility of material space design for ahead. This study suggest an alternative idea for revitalizing design by materials.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3367-3378
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• 2021

The design phase and safety assessment of Generation IV liquid metal-cooled fast reactors calls for the improvement of fuel pin performance codes, in particular the enhancement of their predictive capabilities towards uranium-plutonium mixed oxide fuels and stainless-steel cladding under irradiation in fast reactor environments. To this end, the current capabilities of fuel performance codes must be critically assessed against experimental data from available irradiation experiments. This work is devoted to the assessment of three European fuel performance codes, namely GERMINAL, MACROS and TRANSURANUS, against the irradiation of two fuel pins selected from the SUPERFACT-1 experimental campaign. The pins are characterized by a low enrichment (~ 2 wt.%) of minor actinides (neptunium and americium) in the fuel, and by plutonium content and cladding material in line with design choices envisaged for liquid metal-cooled Generation IV reactor fuels. The predictions of the codes are compared to several experimental measurements, allowing the identification of the current code capabilities in predicting fuel restructuring, cladding deformation, redistribution of actinides and volatile fission products. The integral assessment against experimental data is complemented by a code-to-code benchmark focused on the evolution of quantities of engineering interest over time. The benchmark analysis points out the differences in the code predictions of fuel central temperature, fuel-cladding gap width, cladding outer radius, pin internal pressure and fission gas release and suggests potential modelling development paths towards an improved description of the fuel pin behaviour in fast reactor irradiation conditions.

• Nuclear Engineering and Technology
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• 제19권1호
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• pp.22-33
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• 1987

원자력 발전소에 있어서 정상가동 상태나 이상동작시에 핵연료 피복관의 건전성 확보와 연관하여 피복재의 항복거동은 중요한 문제이다. 급격한 출력상승 상황에서 이산화 우라늄 소결체와 피복관 사이의 노내 조사거동의 차이는 소결체와 피복관 사이에 Contact Pressure를 야기 시킨다. 만일 이 Contact Pressure가 Zircaloy 피복관의 Yield Pressure에 도달하면 피복관에는 영구변형이 일어난다. 이 변형은 원자로의 출력이 정상상태로 회복되더라도 존재하므로 소결체와 피복관 사이의 Gap을 증대시킨다. 이러한 상황을 묘사하기 위해 본 논문에서는 구리 Mandrel과 Zircaloy사이의 열팽창 차이를 이용하는 Mandrel 팽창 실험을 실행했다. 실험 결과 측정된 Zircaloy 피복관의 외경 팽창치와 본 논문에서 유도된 수학적 관계식들을 이용하여 온도에 따른 Zircaloy 피복관의 내부항복압력과 항복응력, 피복재의 항복에 따른 핵연료 소결체와 피복관 사이의 Gap 증대를 구하고, 항복 거동에 따른 온도의 영향을 보기 위해 항복과정의 활성화 에너지를 구했다. 본 실험과 분석에서 얻어진 이들 결과들은 다른 실험 결과들과 상당히 일치하였으며, 이것으로 볼 때 본 논문에서 유도된 관계식들과 Mandrel 팽창 실험이 Zircaloy 피복관의 항복거동과 Gap Expansion 측정에 신뢰성이 있음을 알 수 있었다.

• 한국강구조학회 논문집
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• 제20권5호
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• pp.591-599
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• 2008

내후성강은 자연대기환경에 대하여 무도장 상태로 사용되고, 강재표면에 도막의 역할을 하는 밀실한 녹층은 자연환경에 의해 수년간 에 걸쳐 형성되며, 그 녹 층으로 강재의 부식 진행을 억제한다. 내후성강의 녹상태는 부식환경에 의존한다. 내후성강의 유지관리측면에서 녹상태가 어느정도 인가를 판단하는 것은 중요하다. 따라서, 국내에 있는 건축용 내후성 외장재의 외관조사결과를 바탕으로, 다변량해석에 의한 부식량(녹의 상태)과 외장재 주변환경과의 상관을 분석하고자 한다. 본 논문에서, 외장재의 지리적 데이터, 기상데이터를 이용한 다변량 분석이 수행되었고, 부식 량과 환경요인에 대한 예측식이 제안되었다. 부가적으로, 제안된 부식량 예측식을 이용하여 기호지역(서울, 경기 및 충청)에 적용한 건축용 내후성강 외장재의 부식량 추이를 평가하였다.

• 한국세라믹학회지
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• 제50권6호
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• pp.359-363
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• 2013

Silicon carbide and its composites are being considered as a nuclear fuel cladding material for LWR nuclear reactors because they have a low neutron absorption cross section, low hydrogen production under accident conditions, and high strength at high temperatures. The SiC composite cladding tube considered in this study consists of three layers, monolith CVD SiC - $SiC_f$/SiC composite -monolith CVD SiC. The volume fraction of SiC fiber and surface roughness of the composite layer affect mechanical and corrosion properties of the cladding tube. In this study, various types of SiC fiber preforms with tubular shapes were fabricated by a filament winding method using two types of Tyranno SA3 grade SiC fibers with 800 filaments/yarn and 1600 filaments/yarn. After chemical vapor infiltration of the SiC matrix, the surface roughness and fiber volume fraction were measured. As filament counts were changed from 800 to 1600, the surface roughness increased but the fiber volume fraction decreased. The $SiC_f$/SiC composite with a bamboo-like winding pattern has a smaller surface roughness and a higher fiber volume fraction than that with a zigzag winding pattern.

• Nuclear Engineering and Technology
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• 제49권4호
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• pp.734-743
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• 2017

In order to investigate the swelling behavior and fuel-cladding interaction mechanism of U-10Zr alloy metallic fuel at very low burnup, an irradiation experiment was concisely designed and conducted on the China Mianyang Research Reactor. Two types of irradiation samples were designed for studying free swelling without restraint and the fuel-cladding interaction mechanism. A new bonding material, namely, pure aluminum powder, was used to fill the gap between the fuel slug and sample shell for reducing thermal resistance and allowing the expansion of the fuel slug. In this paper, the concise irradiation rig design is introduced, and the neutronic and thermal-hydraulic analyses, which were carried out mainly using MCNP (Monte Carlo N-Particle) and FLUENT codes, are presented. Out-of-pile tests were conducted prior to irradiation to verify the manufacturing quality and hydraulic performance of the rig. Nondestructive postirradiation examinations using cold neutron radiography technology were conducted to check fuel cladding integrity and swelling behavior. The results of the preliminary examinations confirmed the safety and effectiveness of the design.

• Nuclear Engineering and Technology
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• 제20권4호
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• pp.241-245
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• 1988

핵연료봉의 PCI 파손은 원자로의 운전제한과 밀접한 관계가 있기 때문에, 출력급증 조건에서 핵연료봉의 PCI 파손을 지배하는 파손인자들의 거동을 검토하는 것은 매우 중요하다. 본 연구에서는 피복관에서의 원주방향 응력, 원주방향 변형도, 원주방향 주름 높이, 크립 변형율 및 변형도 에너지등의 파손인자들에 대한 거동특성을 핵연료봉 성능해석용 전산코드인 FEMAXI-IV를 이용하여 출력급증량 및 출력증가율의 운전인자들의 함수로 검토하였다.

• 한국분말야금학회:학술대회논문집
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• 한국분말야금학회 2006년도 Extended Abstracts of 2006 POWDER METALLURGY World Congress Part 1
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• pp.293-294
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• 2006

Laser Powder Deposition (LPD) is a technology capable of modifying a metallic structure by adding the appropriate material to perform a desired function. LPD offers a unique fabrication technique that allows the use of soft (tough) materials as base structures. Through LPD a hard material can be applied to the base material with little thermal input (minimal dilution and heat-affected-zone {HAZ}), thus providing the function of a heat treatment or other surface modifications. These surface modifications have been evaluated through standard wear testing (ASTM G-65), surface hardness (Rc), micro-hardness (vickers), and optical microscopy.