• 한국방사성폐기물학회:학술대회논문집
• /
• 한국방사성폐기물학회 2017년도 추계학술논문요약집
• /
• pp.127-128
• /
• 2017

• Nuclear Engineering and Technology
• /
• 제50권2호
• /
• pp.229-236
• /
• 2018

Accident-tolerant fuels (ATFs) are currently of high interest to researchers in the nuclear industry and in governmental and international organizations. One widely studied accident-tolerant fuel concept is multilayer cladding (also known as coated cladding). This concept is based on a traditional Zr-based alloy (Zircaloy-4, M5, E110, ZIRLO etc.) serving as a substrate. Different protective materials are applied to the substrate surface by various techniques, thus enhancing the accident tolerance of the fuel. This study focuses on the results of testing of Zircaloy-4 coated with pure chromium metal using the cold spray (CS) technique. In comparison with other deposition methods, e.g., Physical vapor deposition (PVD), laser coating, or Chemical vapor deposition techniques (CVD), the CS technique is more cost efficient due to lower energy consumption and high deposition rates, making it more suitable for industry-scale production. The Cr-coated samples were tested at different conditions ($500^{\circ}C$ steam, $1200^{\circ}C$ steam, and Pressurized water reactor (PWR) pressurization test) and were precharacterized and postcharacterized by various techniques, such as scanning electron microscopy, Energy-dispersive X-ray spectroscopy (EDX), or nanoindentation; results are discussed. Results of the steady-state fuel performance simulations using the Bison code predicted the concept's feasibility. It is concluded that CS Cr coating has high potential benefits but requires further optimization and out-of-pile and in-pile testing.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1998년도 춘계학술발표회논문집(2)
• /
• pp.175-180
• /
• 1998

The U-Zr metallic alloy with low swelling HT9 cladding is the candidate for the KALIMER fuel rod. The fuel rod should be able to maintain the structural integrity during its lifetime in the reactor. In a typical metallic fuel rod, load is mainly applied by internal gas pressure, and the deformation is primarily caused by creep of the cladding. The three-dimensional FEM modelling of a fuel rod is important to predict the structural behavior in concept design stage. Using the ANSYS code, the 3-D structure analyses were performed for various configuration, element and loads. It has been shown that the present analysis model properly evaluate the structural integrity of fuel rod. The present analysis results show that the fuel rod is expected to maintain its structural integrity during normal operation.

• Nuclear Engineering and Technology
• /
• 제41권2호
• /
• pp.171-178
• /
• 2009

The rate of delayed hydride cracking (DHC), V, has been measured in cold-worked and stress-relieved Zircaloy-4 fuel cladding using the Pin-Loading Tension technique. At $250^{\circ}C$ the mean value of V from 69 specimens was $3.3({\pm}0.8)x10^{-8}$ m/s while the temperature dependence up to $275^{\circ}C$ was described by Aexp(-Q/RT), where Q is 48.3 kJ/mol. No cracking or cracking at very low rates was observed at higher temperatures. The fracture surface consisted of flat fracture with no striations. The results are compared with previous results on fuel cladding and pressure tubes.

• Nuclear Engineering and Technology
• /
• 제21권4호
• /
• pp.332-340
• /
• 1989

핵연료의 로내 연소거동 분석평가 연구의 일환으로 가압경수로에서 3주기동안 연소한 14$\times$14 사용후 핵연료를 핫셀에서 파괴시험하여 다음과 같은 결과를 얻었다. 1) 고연소 부위의 연료중심에서도 핵연료의 결정립성장은 일어나지 않았다. 2) 연소도 증가에 따라 밀도감소가 일어나 36,000 MWD/MTU 연소도에서는 연료의 밀도가 94.4% TD까지 감소하였다. 3) 피복관의 평균 산화층두께는 연료봉의 중간 및 하부부위에서는 10$\mu$m이하였으나 상부부 위 에서는 급격하게 20$\mu$m이 상으로 증가되었다. 4) 피복관의 수소화물 생성량은 피복관의 산화물 생성량가 연동되어 연료봉 하부보다는 상부에서는 생성량이 많았다.

• Nuclear Engineering and Technology
• /
• 제50권5호
• /
• pp.724-730
• /
• 2018

Chromium and chromium nitride were selected as potential barriers to prevent fuel-clad chemical interaction (FCCI) between the cladding and the fuel material. In this study, ferritic/martensitic HT-9 steel and misch metal were used to simulate the reaction between the cladding and fuel fission product, respectively. Radio frequency magnetron sputtering was used to deposit Cr and CrN films onto the cladding, and the gas flow rates of argon and nitrogen were fixed at certain values for each sample to control the deposition rate and the crystal structure of the films. The samples were heated for 24 h at 933 K through the diffusion couple test, and considerable amount of interdiffusion (max. thickness: $550{\mu}m$) occurred at the interface between HT-9 and misch metal when the argon and nitrogen were used individually. The elemental contents of misch metal were detected at the HT-9 through energy dispersive X-ray spectroscopy due to the interdiffusion. However, the specimens that were sputtered by mixed gases (Ar and $N_2$) exhibited excellent resistance to FCCI. The thickness of these CrN films were only $4{\mu}m$, but these films effectively prevented the FCCI due to their high adhesion strength (frictional force ${\geq}1,200{\mu}m$) and dense columnar microstructures.

• Nuclear Engineering and Technology
• /
• 제53권4호
• /
• pp.1236-1249
• /
• 2021

Although there are still controversial opinions and uncertainty on application of SiC_f/SiC composite cladding as next-generation cladding material for its great oxidation resistance in high temperature steam environment and other outstanding advantages, it cannot deny that SiC_f/SiC cladding is a potential accident tolerant fuel (ATF) cladding with high research priority and still in the engineering design stage for now. However, considering its disadvantages, such as low irradiated thermal conductivity, ductility that barely not exist, further evaluations of its in-pile behaviors are still necessary. Based on the self-developed code we recently updated, relevant thermohydraulic and mechanical models in FROBA-ATF were applied to simulate the cladding behaviors under normal and accident conditions in this paper. Even through steady-state performance analysis revealed that this kind of cladding material could greatly reduce the oxidation thickness, the thermal performance of UO₂-SiC was poor due to its low inpile thermal conductivity and creep rate. Besides, the risk of failure exists when reactor power decreased. With geometry optimization and dopant addition in pellets, the steady-state performance of UO₂-SiC was enhanced and the failure risk was reduced. The thermal and mechanical performance of the improved UO₂-SiC was further evaluated under Loss of coolant accident (LOCA) and Reactivity Initiated Accident (RIA) conditions. Transient results showed that the optimized ATF had better thermal performance, lower cladding hoop stress, and could provide more coping time under accident conditions.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2004년도 추계학술발표회 발표논문집
• /
• pp.1026-1027
• /
• 2004

• Nuclear Engineering and Technology
• /
• 제48권5호
• /
• pp.1096-1108
• /
• 2016

Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U-transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

• Nuclear Engineering and Technology
• /
• 제25권4호
• /
• pp.570-587
• /
• 1993

최근 전 세계적으로 내부2차수소 침투에 의한 것으로 보이는 핵연료의 심각한 파손이 잇달아 보고되었다. 본 논문에서는 결함 핵연료에서의 내부 수소 침투 현상이 개괄적으로 고찰된다. 먼저 핵연료의 피복관으로 사용되는 질코늄 합금의 개발사와 그 질코늄 합금이 사용되는 원자로 내의 운전조건이 소개되고 산화 질코늄 막에서의 수소의 투과성, 질코늄과 질코늄 합금에서의 수소의 최종 용해도와 침전도등 질코늄의 수소 침투에 관련된 기본 사항과 수소 침투가 기계적 강도에 미치는 악 영향등이 고찰된다. 결함 핵연료봉 내부에서 발생되는 수소의 대량 내부 침투의 메카니즘이 관찰된 제 현상을 중심으로 정성적으로 설명되고 이러한 수소의 대량 침투에 의한 핵연료치 파손심화를 줄이기 위한 노력의 일환으로 제시된 정량적 모델이 간단히 언급되고 이러한 정량적인 모델의 심도있는 개발을 위해 필요한 자료와 추후의 연구 내용이 설명된다.